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

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

Assessment of FBR MONJU accident management reliability in causing reactor trips

International Nuclear Information System (INIS)

Sotsu, Masutake; Kurisaka, Kenichi

2010-01-01

This paper describes a method and application of quantitatively evaluating Accident Management (AM) reliability upon a reactor trip failure for the MONJU fast breeder reactor using a PSA technique. The present method comprises an allowable time estimation that is based on plant transient response analysis using the Super-COPD code that was developed for use in best estimates of the plant dynamics of MONJU and in estimating failure probability of operator's actions in AMs within the allowable time based on time records obtained from simulator training. Application of this method to MONJU resulted in the estimation that the allowable time for an unprotected loss-of-heat sink event would be more than the longest observed time of 326 s. The corresponding operation failure probability would be less than 0.1 even after taking the uncertainty into consideration. Combining this with a level 1 PSA revealed that the total frequency of core damage accompanying a reactor trip failure at MONJU could be decreased by at least 50 percent due to the reactor trip AM. (author)

Analysis of reactor trips originating in balance of plant systems

International Nuclear Information System (INIS)

Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W.

1990-09-01

This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

Reducing scram frequency by modifying/eliminating steam generator low-low level reactor trip setpoint for Maanshan nuclear power plant

International Nuclear Information System (INIS)

Yuann, R.Y.; Chiang, S.C.; Hsiue, J.K.; Chen, P.C.

1987-01-01

The feasibility of modification/elimination of steam generator low-low level reactor trip setpoint is evaluated by using RETRAN-02 code for the purpose of reducing scram frequency in Maanshan 3-loop pressurized water reactor. The ANS Condition II event loss of normal feedwater and condition IV event feedwater system line break are the basis for steam generator low-low level reactor trip setpoint sensitivity analysis, including various initial reactor power levels, reactivity feedback coefficients, and system functions assumptions etc., have been performed for the two basis events with steam generator low-low level reactor trip setpoint at 0% narrow range and without this trip respectively. The feasibility of modifying/eliminating current steam generator low-low level reactor trip setpoint is then determined based on whether the analysis results meet with the ANS Condition II and IV acceptance criteria or not

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.

Task types and error types involved in the human-related unplanned reactor trip events

International Nuclear Information System (INIS)

Kim, Jae Whan; Park, Jin Kyun

2008-01-01

In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1%), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed

Task types and error types involved in the human-related unplanned reactor trip events

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Park, Jin Kyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-12-15

In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1%), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed.

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

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.)

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).

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)

Automatic trip and mode detection with MoveSmarter: first results from the Dutch Mobile Mobility Panel

NARCIS (Netherlands)

Geurs, Karst Teunis; Thomas, Tom; Bijlsma, Marcel; Douhou, Salima

2015-01-01

This paper describes the performance of a smartphone app called MoveSmarter to automatically detect departure and arrival times, trip origins and destinations, transport modes, and travel purposes. The app is used in a three-year smartphone-based prompted-recall panel survey in which about 600

Basic Characteristics of Human Erroneous Actions during Test and Maintenance Activities Leading to Unplanned Reactor Trips

International Nuclear Information System (INIS)

Kim, Jae Whan; Park, Jin Kyun

2010-01-01

Test and maintenance (T and M) activities of nuclear power plants are essential for sustaining the safety of a power plant and maintaining the reliability of plant systems and components. However, the potential of human errors during T and M activities has also the potential to induce unplanned reactor trips or power derate or making safety-related systems unavailable. According to the major incident/accident reports of nuclear power plants in Korea, contribution of human errors takes up about 20% of the total events. The previous study presents that most of human-related unplanned reactor trip events during normal power operation are associated with T and M activities (63%), which are comprised of plant maintenance activities such as a 'periodic preventive maintenance (PPM)', a 'planned maintenance (PM)' and a 'corrective maintenance (CM)'. This means that T and M activities should be a major subject for reducing the frequency of human-related unplanned reactor trips. This paper aims to introduce basic characteristics of human erroneous actions involved in the test and maintenance-induced unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants. The basic characteristics are described by dividing human erroneous actions into planning-based errors and execution-based errors. For the events associated with planning failures, they are, firstly, classified according to existence of the work procedure and then described for what aspects of the procedure or work plan have deficiency or problem. On the other hand, for the events associated with execution failures, they are described from the aspect of external error modes

Development of an automatic reactor inspection system

International Nuclear Information System (INIS)

Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

2002-02-01

Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

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

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)

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

Reactor component automatic grapple

International Nuclear Information System (INIS)

Greenaway, P.R.

1982-01-01

A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment. (author)

Automatic control system in the reactor peggy

International Nuclear Information System (INIS)

Bertrand, J.; Mourchon, R.; Da Costa, D.; Desandre-Navarre, Ch.

1967-01-01

The equipment makes it possible for the reactor to attain a given power automatically and for the power to be maintained around this level. The principle of its operation consists in the changing from one power to another, at constant period, by means of a programmer transforming a power-step request into a voltage variation which is linear with time and which represents the logarithm of the required power. The real power is compared continuously with the required power. Stabilization occurs automatically as soon as the difference between the reactor power and the required power diminishes to a few per cent. (authors) [fr

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)

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)

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

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)

Automatic power control for a pressurized water reactor

International Nuclear Information System (INIS)

Hah, Yung Joon

1994-02-01

During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

Doosan Experience on I and C Upgrade for Operating NPPs: Control Rod Control System and Automatic Seismic Trip System

International Nuclear Information System (INIS)

Nam, C.H.; Kim, K.H.; Lee, D.H.

2012-01-01

This paper describes DHIC's experience on upgrading 3 coil type control rod control system(CRCS), 4 coil type control element drive mechanism control system(CEDMCS) and automatic seismic trip system(ASTS). Common main feature of the above systems are full duplex system to prevent unwanted trip and mis-operation. 5 CRCS and CEDMCS have been supplied to Kori 1,2, Ulchin 1,2 and Younggwang 3 since 2010 and 7 CEDMCS are contracted to supply Korea Hydro and Nuclear Power Co.(KHNP) site. Also 16 ASTS are supplied and 12 ASTS will be supplied to operating and new NPPs within 3 years. (author)

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

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)

Fuzzy algorithm for an automatic reactor power control in a PWR

International Nuclear Information System (INIS)

Hah, Yung Joon; Song, In Ho; Yu, Sung Sik; Choi, Jung In; Lee, Byong Whi

1994-01-01

A fuzzy algorithm is presented for automatic reactor power control in a pressurized water reactor. Automatic power shape control is complicated by the use of control rods because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability for the load - follow operation including frequency control. In an attempt to achieve automatic power shape control without any design modification of the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multi - input multi - output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to the Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of the pressurized water reactor during the load - follow operation

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)

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)

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

Automatic start-up system of nuclear reactor based on sequence control technology

International Nuclear Information System (INIS)

Zhang Yao; Zhang Dafa; Peng Huaqing

2009-01-01

A conceptive design of an automatic start-up system based on the sequence control for the nuclear reactors is given in this paper, so as to solve the problems during the start-up process, such as the long operation time, low automatic control level and high accident rate. The start-up process and its requirements are analyzed in detail at first. Then,the principle, the architecture, the key technologies of the automatic start-up system of nuclear reactors are designed and discussed. With the designed system, the automatic start-up of the nuclear reactor can be realized,the work load of the operator can be reduced,and the safety and efficiency of the nuclear power plant during its start-up can be improved. (authors)

Simulation of the TREAT-Upgrade Automatic Reactor Control System

International Nuclear Information System (INIS)

Lipinski, W.C.; Kirsch, L.W.; Valente, A.D.

1984-01-01

This paper describes the design of the Automatic Reactor Control System (ARCS) for the Transient Reactor Test Facility (TREAT) Upgrade. A simulation was used to facilitate the ARCS design and to completely test and verify its operation before installation at the TREAT facility

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

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

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)

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)

Analysis of the oscillation causes in automatic controller of reactor power

International Nuclear Information System (INIS)

Aleksakov, A.N.; Nikolaev, E.V.; Podlazov, L.N.

1991-01-01

Conditions for occurence of oscillations in automatic controller of reactor power are determined. Graphic-analytical method for calculating the stability of non-linear system, which enables one to reveal the most important factors determining the stability, is used. The practical results of the analysis are obtained for the system of local automatic comtrollers, used in the RBMK reactors. A simple method providing for the required stability margin, is suggested

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

Model study of an automatic controller of the IBR-2 pulsed reactor

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Popov, A.K.

2007-01-01

For calculation of power transients in the IBR-2 reactor a special mathematical model of dynamics taking into account the discontinuous jump of reactivity by an automatic controller with the step motor is created. In the model the nonlinear dependence of the energy of power pulse on the reactivity and the influence of warming up of the reactor on the reactivity by means of introduction of a nonlinear feedback 'power-pulse energy - reactivity' are taken into account. With the help of the model the transients of relative deviation of power-pulse energy are calculated at various (random, mixed and regular) reactivity disturbances at the reactor mean power 1.475 MW. It is shown that to improve the quality of processes the choice of such regular values of parameters of the automatic controller is expedient, at which the least effect of smoothing of a signal acting on an automatic controller and the least speed of an automatic controller are provided, and the reduction of efficiency of one step of the automatic controller and introduction of a five-percent dead space are also expedient

Automatic control of the water level of steam generators from 0% to 100% of the load

International Nuclear Information System (INIS)

Hocepied, R.; Debelle, J.; Timmermans, A.; Lams, J.-L.; Baeyens, R.; Eussen, G.; Bassem, G.

1978-01-01

The water level of a steam generator is hard to control manually and it is practically impossible for a human operator to react correctly to every important perturbation. These phenomena are further accentuated during the start-up at low load and at low feedwater temperature. The control schemes traditionally provided do not permit satisfactory automatic level control during all operating circumstances. Adaptions of the control system permit all the problems encountered to be solved: automatic control of the level in the steam generators is possible from 0% to 100% of the load and also when large-scale perturbations occur. Such a result has been obtained by use of systematic methods for the analysis of the steam generator's behaviour. These methods have also been used to verify the performance of the control system. The control system installed at the Doel nuclear power station prevents most of the reactor or turbine trip-outs caused by level deviations occurring during start-up and low-load operation. It also minimizes the effects on the unit of incidents such as tripping the unit on house load, safety tripping, fast run-back on reduced load, etc. The principles used are applicable to the control of steam generators of all pressurized water reactor power stations. (author)

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

Stop valve with automatic control and locking for nuclear reactors

International Nuclear Information System (INIS)

Chung, D.K.

1980-01-01

This invention generally concerns an automatic control and locking stop valve. Specifically it relates to the use of such a valve in a nuclear reactor of the type containing absorber elements supported by a fluid and intended for stopping the reactor in complete safety [fr

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)

Experience with automatic reactor control at EBR-II

International Nuclear Information System (INIS)

Lehto, W.K.; Larson, H.A.; Christensen, L.J.

1985-01-01

Satisfactory operation of the ACRDS has extended the capabilities of EBR-II to a transient test facility, achieving automatic transient control. Test assemblies can now be irradiated in transient conditions overlapping the slower transient capability of the TREAT reactor

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

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)

An investigation on unintended reactor trip events in terms of human error hazards of Korean nuclear power plants

International Nuclear Information System (INIS)

Kim, Sa Kil; Lee, Yong Hee; Jang, Tong Il; Oh, Yeon Ju; Shin, Kwang Hyeon

2014-01-01

Highlights: • A methodology to identify human error hazards has been established. • The proposed methodology is a preventive approach to identify not only human error causes but also its hazards. • Using the HFACS framework we tried to find out not causations but all of the hazards and relationships among them. • We determined countermeasures against human errors through dealing with latent factors such as organizational influences. - Abstract: A new approach for finding the hazards of human errors, and not just their causes, in the nuclear industry is currently required. This is because finding causes of human errors is really impossible owing to the multiplicity of causes in each case. Thus, this study aims at identifying the relationships among human error hazards and determining the strategies for preventing human error events by means of a reanalysis of the reactor trip events in Korea NPPs. We investigated human errors to find latent factors such as decisions and conditions in all of the unintended reactor trip events during the last dozen years. In this study, we applied the HFACS (Human Factors Analysis and Classification System), which is a commonly utilized tool for investigating human contributions to aviation accidents under a widespread evaluation scheme. Using the HFACS framework, we tried to find out not the causations but all of the hazards and their relationships in terms of organizational factors. Through the trial, we proposed not only meaningful frequencies of each hazards also correlations of them. Also, considering the correlations of each hazards, we suggested useful strategies to prevent human error event. A method to investigate unintended nuclear reactor trips by human errors and the results will be discussed in more detail

Design of Simulink module for dynamic reactivity simulation of marine reactor automatic control rod

International Nuclear Information System (INIS)

Chen Zhiyun; Luo Lei; Chen Wenzhen; Gui Xuewen

2010-01-01

The power of marine reactor varies frequently and acutely, which induces the frequent and acute adjustment of the automatic control rod. According to the characteristics of marine reactor and the problem of improper control rod reactivity insertion in previous literatures, the Simulink module for dynamic reactivity simulation of automatic control rod was designed and adopted as a sub-module of Simulink program for the fast calculation of the physical and thermal parameters of marine reactor. A typical dynamic process of the marine reactor was used as the benchmark, which indicates that the designed Simulink module is capable of the dynamic simulation of automatic control rod position and reactivity, and is adequate to the fast calculation of physic and thermal parameters. The Simulink module is of significant meaning to the simulation of the dynamic process of marine reactor and the fast calculation of the operating parameters. (authors)

Use of reactivity constraints for the automatic control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.; Lanning, D.D.; Ray, A.

1985-01-01

A theoretical framework for the automatic control of reactor power has been developed and experimentally evaluated on the 5 MWt Research Reactor that is operated by the Massachusetts Institute of Technology. The controller functions by restricting the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally designated as ''feasibility of control''. It has been shown experimentally that maintenance of feasibility of control is a sufficient condition for the automatic control of reactor power. This research should be of value in the design of closed-loop controllers, in the creation of reactivity displays, in the provision of guidance to operators regarding the timing of reactivity changes, and as an experimental envelope within which alternate control strategies can be evaluated

Reactor power automatically controlling method and device for BWR type reactor

International Nuclear Information System (INIS)

Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

1997-01-01

For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

Reactor protection system with automatic self-testing and diagnostic

International Nuclear Information System (INIS)

Gaubatz, D.C.

1996-01-01

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

Energy Technology Data Exchange (ETDEWEB)

Tzang, Y.-C., E-mail: yctzang@aec.gov.t [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China); Chiang, R.-F.; Ferng, Y.-M.; Pei, B.-S. [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China)

2009-12-15

In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the DELTACPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

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

Automatic Control of Reactor Temperature and Power Distribution for a Daily Load following Operation

Energy Technology Data Exchange (ETDEWEB)

Yu, Keuk Jong; Kim, Han Gon [Korea Hydro and Nuclear Power Institute, Daejeon (Korea, Republic of)

2010-10-15

An automatic control method of reactor power and power distribution was developed for a daily load following operation of APR1400. This method used a model predictive control (MPC) methodology having second-order plant data. And it utilized a reactor power ratio and axial shape index as control variables. However, the reactor regulating system of APR1400 is operated by the difference between the average temperature of the reactor core and the reference temperature, which is proportional to the turbine load. Thus, this paper reports on the model predictive control methodology using fourth-order plant data and a reactor temperature instead of the reactor power shape. The purpose of this study is to develop a revised automatic controller and analyze the behavior of the nuclear reactor temperature (Tavg) and the axial shape index (ASI) using the MPC method during a daily load following operation

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)

Analysis of Task Types and Error Types of the Human Actions Involved in the Human-related Unplanned Reactor Trip Events

International Nuclear Information System (INIS)

Kim, Jae Whan; Park, Jin Kyun; Jung, Won Dea

2008-02-01

This report provides the task types and error types involved in the unplanned reactor trip events that have occurred during 1986 - 2006. The events that were caused by the secondary system of the nuclear power plants amount to 67 %, and the remaining 33 % was by the primary system. The contribution of the activities of the plant personnel was identified as the following order: corrective maintenance (25.7 %), planned maintenance (22.8 %), planned operation (19.8 %), periodic preventive maintenance (14.9 %), response to a transient (9.9 %), and design/manufacturing/installation (9.9%). According to the analysis of error modes, the error modes such as control failure (22.2 %), wrong object (18.5 %), omission (14.8 %), wrong action (11.1 %), and inadequate (8.3 %) take up about 75 % of all the unplanned trip events. The analysis of the cognitive functions involved showed that the planning function makes the highest contribution to the human actions leading to unplanned reactor trips, and it is followed by the observation function (23.4%), the execution function (17.8 %), and the interpretation function (10.3 %). The results of this report are to be used as important bases for development of the error reduction measures or development of the error mode prediction system for the test and maintenance tasks in nuclear power plants

Analysis of Task Types and Error Types of the Human Actions Involved in the Human-related Unplanned Reactor Trip Events

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Park, Jin Kyun; Jung, Won Dea

2008-02-15

This report provides the task types and error types involved in the unplanned reactor trip events that have occurred during 1986 - 2006. The events that were caused by the secondary system of the nuclear power plants amount to 67 %, and the remaining 33 % was by the primary system. The contribution of the activities of the plant personnel was identified as the following order: corrective maintenance (25.7 %), planned maintenance (22.8 %), planned operation (19.8 %), periodic preventive maintenance (14.9 %), response to a transient (9.9 %), and design/manufacturing/installation (9.9%). According to the analysis of error modes, the error modes such as control failure (22.2 %), wrong object (18.5 %), omission (14.8 %), wrong action (11.1 %), and inadequate (8.3 %) take up about 75 % of all the unplanned trip events. The analysis of the cognitive functions involved showed that the planning function makes the highest contribution to the human actions leading to unplanned reactor trips, and it is followed by the observation function (23.4%), the execution function (17.8 %), and the interpretation function (10.3 %). The results of this report are to be used as important bases for development of the error reduction measures or development of the error mode prediction system for the test and maintenance tasks in nuclear power plants.

Microcontroller based automatic liquid poison addition control system

International Nuclear Information System (INIS)

Kapatral, R.S.; Ananthakrishnan, T.S.; Pansare, M.G.

1989-01-01

Microcontrollers are finding increasing applications in instrumentation where complex digital circuits can be substituted by a compact and simple circuit, thus enhancing the reliability. In addition to this, intelligence and flexibility can be incorporated. For applications not requiring large amount of read/write memory (RAM), microcontrollers are ideally suited since they contain programmable memory (Eprom), parallel input/output lines, data memory, programmable timers and serial interface ports in one chip. This paper describes the design of automatic liquid poison addition control system (ALPAS) using intel's 8 bit microcontroller 8751, which is used to generate complex timing control sequence signals for liquid poison addition to the moderator in a nuclear reactor. ALPAS monitors digital inputs coming from protection system and regulating system of a nuclear reactor and provides control signals for liquid poison addition for long term safe shutdown of the reactor after reactor trip and helps the regulating system to reduce the power of the reactor during operation. Special hardware and software features have been incorporated to improve performance and fault detection. (author)

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.

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

Automatic optimized reload and depletion method for a pressurized water reactor

International Nuclear Information System (INIS)

Ahn, D.H.; Levene, S.H.

1985-01-01

A new method has been developed to automatically reload and deplete a pressurized water reactor (PWR) so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: (a) determination of the minimum fuel requirement for an equivalent three-region core model, (b) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the reload cost, (c) optimal placement of fuel assemblies to conserve regionwise optimal conditions, and (d) optimal control through poison management to deplete individual fuel assemblies to maximize end-of-cycle k /SUB eff/ . The new method differs from previous methods in that the optimization process automatically performs all tasks required to reload and deplete a PWR. In addition, the previous work that developed optimization methods principally for the initial reactor cycle was modified to handle subsequent cycles with fuel assemblies having burnup at beginning of cycle. Application of the method to the fourth reactor cycle at Three Mile Island Unit 1 has shown that both the enrichment and the number of fresh reload fuel assemblies can be decreased and fully amortized fuel assemblies can be reused to minimize the fuel cost of the reactor

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

Device for the nuclear reactor automatic start-up and power control

International Nuclear Information System (INIS)

Nikiforov, B.N.; Volkov, A.V.; Ogon'kov, A.I.

1978-01-01

A description and flowsheet of a reactor start-up and power-control automatic device containing no nonlinear elements with a relay characteristic are presented. The device consists of two independent channels for measuring the physical power and time (period) constant of the reactor. Requirements for the device are considered, based on the condition of a minimum permissible number of a servomechanism operations due to fluctuations of an input signal which appear because of the statistical nature of processes taking place in the reactor. It is noted that the threshold amplifier used in the device allows a considerable decrease of the reactor start-up time

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)

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.)

The automatic programming for safety-critical software in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Eom, Heung Seop; Choi, You Rark

1998-06-01

We defined the Korean unique safety-critical software development methodology by modifying Dr. Harel`s statechart-based on formal methods in order to digitalized the reactor protection system. It is suggested software requirement specification guideline to specify design specification which is basis for requirement specification and automatic programming by the caused by shutdown parameter logic of the steam generator water level for Wolsung 2/3/4 unit SDS no.1 and simulated it by binding the Graphic User Interface (GUI). We generated the K and R C code automatically by utilizing the Statemate MAGNUM Sharpshooter/C code generator. Auto-generated K and R C code is machine independent code and has high productivity, quality and provability. The following are the summaries of major research and development. - Set up the Korean unique safety-critical software development methodology - Developed software requirement specification guidelines - Developed software design specification guidelines - Reactor trip modeling for steam generator waster level Wolsung 2/3/4 SDS no. 1 shutdown parameter logic - Graphic panel binding with GUI. (author). 20 refs., 12 tabs., 15 figs

The automatic programming for safety-critical software in nuclear power plants

International Nuclear Information System (INIS)

Kim, Jang Yeol; Eom, Heung Seop; Choi, You Rark

1998-06-01

We defined the Korean unique safety-critical software development methodology by modifying Dr. Harel's statechart-based on formal methods in order to digitalized the reactor protection system. It is suggested software requirement specification guideline to specify design specification which is basis for requirement specification and automatic programming by the caused by shutdown parameter logic of the steam generator water level for Wolsung 2/3/4 unit SDS no.1 and simulated it by binding the Graphic User Interface (GUI). We generated the K and R C code automatically by utilizing the Statemate MAGNUM Sharpshooter/C code generator. Auto-generated K and R C code is machine independent code and has high productivity, quality and provability. The following are the summaries of major research and development. - Set up the Korean unique safety-critical software development methodology - Developed software requirement specification guidelines - Developed software design specification guidelines - Reactor trip modeling for steam generator waster level Wolsung 2/3/4 SDS no. 1 shutdown parameter logic - Graphic panel binding with GUI. (author). 20 refs., 12 tabs., 15 figs

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.

Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

International Nuclear Information System (INIS)

Severin, V.P.

2007-01-01

The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

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)

Voltage Sag Compensator for CAR and SOR of HANARO

International Nuclear Information System (INIS)

Kim, Hyung-Kyoo; Jung, Hoan-Sung; Wu, Jong-Sup

2007-01-01

HANARO is designed so as to be tripped automatically by insertion of control absorber rods(CAR) and shut-off rods(SOR) and the process systems, such as primary cooling system, secondary cooling system and reflector cooling system, etc., stop whenever the off-site power failure occurs, the reactor trips automatically. When voltage sag or momentary interruption occurs, the process systems are in operation but the reactor has an unwanted trip by insertion of CARs and SORs. We installed the voltage sag compensator on the power supply for CARs and SORs so as to prevent a nuisance trip. The compensated time is decided not to exceed 1 sec in consideration of reactor safety. This paper is concerned with the impact of the momentary interruption on the reactor and the effect of the voltage sag compensator

76 FR 52699 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Science.gov (United States)

2011-08-23

... proposes to credit the automatic trip of the main turbine upon the initiation of a manual reactor trip for... credited manual reactor trip action that is part of the current licensing basis. Considerable defense-in... more robust AREVA Advanced W17 high thermal performance (HTP) fuel at Sequoyah Nuclear Plant (SQN...

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

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

Development of a protection system for research reactor based in Field Programmable Gate Array - FPGA

International Nuclear Information System (INIS)

Martins, Roque Hudson da Silva

2016-01-01

This study presents a implementation purpose of a protection system for research nuclear reactors by using a programed device FPGA (Field Programmable Gate Array). As well as logic protection method involved on an automatic shutdown (TRIP) of a reactor, that ensure the security on such systems. These new control and operation mechanics are developed to guarantee that the security limits of a power plant are not exceeded, these mechanics can work isolated or in groups to safe guard the security levels. For this implementation to be completed, there will be presented the main aspects and concepts referred to protection systems, mostly about research nuclear reactors, with some applications terms exposed. The system proposed at this paper was developed following the VHDL (Very High Speed Integrated Circuits) hardware describing language, and the Modelsim software from Altera Software to program the automatic turning off routines, and hypothetical simulations for such. The results show that for every software application for supporting nuclear reactors, like security devices, they have to meet the IEC 60880 criteria. This paper have great importance, seeing that nuclear reactor security systems, are a basic element for ensure the reactor security. (author)

The Automatic Test Features of the IDiPS Reactor Protection System

International Nuclear Information System (INIS)

Hur, Seop; Kim, Dong-Hoon; Hwang, In-Koo; Lee, Cheol-Kwon; Lee, Dong-Young

2007-01-01

The reactor protection system (RPS) is designed to minimize a propagation of abnormal or accident conditions of nuclear power plants. A digital RPS (Integrated Digital Protection System (IDiPS) RPS) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. To make good use of the advantages of the digital technology, it is necessary to improve the reliability and availability of a system through automatic test features including an on-line testing, a self-diagnostics, an auto calibration, etc. This paper summarizes the system test strategy and the automatic test features of the IDiPS RPS

Investigations on human error hazards in recent unintended trip events of Korean nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Sa Kil; Jang, Tong Il; Lee, Yong Hee; Shin, Kwang Hyeon [KAERI, Daejeon (Korea, Republic of)

2012-10-15

According to the Operational Performance Information System (OPIS) which has been operated to improve the public understanding by the KINS (Korea Institute of Nuclear Safety), unintended trip events by mainly human errors counted up to 38 cases (18.7%) from 2000 to 2011. Although the Nuclear Power Plant (NPP) industry in Korea has been making efforts to reduce the human errors which have largely contributed to trip events, the human error rate might keep increasing. Interestingly, digital based I and C systems is the one of the reduction factors of unintended reactor trips. Human errors, however, have occurred due to the digital based I and C systems because those systems require new or changed behaviors to the NPP operators. Therefore, it is necessary that the investigations of human errors consider a new methodology to find not only tangible behavior but also intangible behavior such as organizational behaviors. In this study we investigated human errors to find latent factors such as decisions and conditions in the all of the unintended reactor trip events during last dozen years. To find them, we applied the HFACS (Human Factors Analysis and Classification System) which is a commonly utilized tool for investigating human contributions to aviation accidents under a widespread evaluation scheme. The objective of this study is to find latent factors behind of human errors in nuclear reactor trip events. Therefore, a method to investigate unintended trip events by human errors and the results will be discussed in more detail.

Investigations on human error hazards in recent unintended trip events of Korean nuclear power plants

International Nuclear Information System (INIS)

Kim, Sa Kil; Jang, Tong Il; Lee, Yong Hee; Shin, Kwang Hyeon

2012-01-01

According to the Operational Performance Information System (OPIS) which has been operated to improve the public understanding by the KINS (Korea Institute of Nuclear Safety), unintended trip events by mainly human errors counted up to 38 cases (18.7%) from 2000 to 2011. Although the Nuclear Power Plant (NPP) industry in Korea has been making efforts to reduce the human errors which have largely contributed to trip events, the human error rate might keep increasing. Interestingly, digital based I and C systems is the one of the reduction factors of unintended reactor trips. Human errors, however, have occurred due to the digital based I and C systems because those systems require new or changed behaviors to the NPP operators. Therefore, it is necessary that the investigations of human errors consider a new methodology to find not only tangible behavior but also intangible behavior such as organizational behaviors. In this study we investigated human errors to find latent factors such as decisions and conditions in the all of the unintended reactor trip events during last dozen years. To find them, we applied the HFACS (Human Factors Analysis and Classification System) which is a commonly utilized tool for investigating human contributions to aviation accidents under a widespread evaluation scheme. The objective of this study is to find latent factors behind of human errors in nuclear reactor trip events. Therefore, a method to investigate unintended trip events by human errors and the results will be discussed in more detail

IE Information Notice No. 85-18, Supplement 1: Failures of undervoltage output circuit boards in the Westinghouse-designed solid state protection system

International Nuclear Information System (INIS)

Rossi, C.E.

1992-01-01

The US Nuclear Regulatory Commission (NRC) is issuing this information notice supplement to alert addressees to continuing problems associated with the undervoltage (UV) output circuit boards (driver cards) in the solid state protection system (SSPS) designed by the Westinghouse Electric Corporation (Westinghouse). On June 3, 1991, the Shearon Harris Nuclear Power Plant, Unit 1, (Harris) experienced an automatic reactor trip from 100 percent power on a spurious low reactor coolant system loop flow signal. The signal was generated as a result of a surveillance test being performed on one of three loop flow transmitters. The licensee attributed the spurious signal to both procedural inadequacies and personnel error. A control room operator verified that all control rods had fully inserted following the trip signal and that reactor power was properly decreasing. However, about 22 seconds after the automatic trip signal was generated, operators discovered that the ''A'' reactor trip breaker (RTB) had not opened. The RTB was manually opened using the reactor trip switch on the main control board. Subsequent analyses are discussed

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

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

Automatic Trip Detection with the Dutch Mobile Mobility Panel: Towards Reliable Multiple-Week Trip Registration for Large Samples

NARCIS (Netherlands)

Thomas, Tom; Geurs, Karst T.; Koolwaaij, Johan; Bijlsma, Marcel E.

2018-01-01

This paper examines the accuracy of trip and mode choice detection of the last wave of the Dutch Mobile Mobility Panel, a large-scale three-year, smartphone-based travel survey. Departure and arrival times, origins, destinations, modes, and travel purposes were recorded during a four week period in

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)

An automatic regulating control system for a graphite moderated reactor using digital techniques

International Nuclear Information System (INIS)

Carvalho Goncalves Junior, J. de.

1989-01-01

The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

Study of a new automatic reactor power control for the TRIGA Mark II reactor at University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Borio Di Tigliole, A.; Magrotti, G. [Laboratorio Energia Nucleare Applicata (L.E.N.A.), University of Pavia, Via Aselli 41, 27100 (Italy); Cammi, A.; Memoli, V. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division (CeSNEF), Via Ponzio 34/3, 20133 Milano (Italy); Gadan, M. A. [Instrumentation and Control Department, National Atomic Energy Comission of Argentina, University of Pavia (Italy)

2009-07-01

The installation of a new Instrumentation and Control (IC) system for the TRIGA Mark-II reactor at University of Pavia has recently been completed in order to assure a safe and continuous reactor operation for the future. The intervention involved nearly the whole IC system and required a channel-by-channel component substitution. One of the most sensitive part of the intervention concerned the Automatic Reactor Power Controller (ARPC) which permits to keep the reactor at an operator-selected power level acting on the control rod devoted to the fine regulation of system reactivity. This controller installed can be set up using different control logics: currently the system is working in relay mode. The main goal of the work presented in this paper is to set up a Proportional-Integral-Derivative (PID) configuration of the new controller installed on the TRIGA reactor of Pavia so as to optimize the response to system perturbations. The analysis have shown that a continuous PID offers generally better results than the relay mode which causes power oscillations with an amplitude of 3% of the nominal power

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.)

Installation of the sag compensator for HANARO

International Nuclear Information System (INIS)

Kim, Hyungkyoo; Jung, Hoansung; Lim, Incheol; Ahn, Gukhoon

2008-01-01

Electric power is essential for all industrial plants and also for nuclear facilities. HANARO is a research reactor which produces a 30 MW thermal power. HANARO is designed to be tripped automatically when interruptions or some extent of sags occur. HANARO has the reactor regulation system(RRS) and reactor protection system(RPS). HANARO is designed so as to be tripped automatically by insertion of control absorber rods(CAR) and shut-off rods(SOR). When voltage sag or momentary interruption occurs, the reactor has an unwanted trip by insertion of CARs and SORs even though the process systems are still in operation. HANARO was experienced in a nuisance trip as often as the unexpected voltage sag and/or momentary interruption occurs. We installed the voltage sag compensator on the power supply for CARs and SORs so as to prevent an unwanted trip. We undertook voltage sag assessment of the AC coil contactor which is a component of the power supply unit for the SORs. The compensation time is determined to be less than 1 sec in consideration of the reactor safety. This paper is concerned with the impact of the momentary interruption on the reactor and the effect of the voltage sag compensator. (author)

Installation of the sag compensator for HANARO

International Nuclear Information System (INIS)

Kim, H. K.; Jung, H. S.; Ahn, G. H.; Lim, I. C.

2008-01-01

Electric power is essential for all industrial plants and also for nuclear facilities. HANARO is a research reactor which produces a 30MW thermal power. HANARO is designed to be tripped automatically when interruptions or some extents of sags occur. HANARO has the reactor regulation system (RRS) and reactor protection system (RPS). HANARO is designed so as to tripped automatically by insertion of control absorber rods (CAR) and shut-off rods (SOR). When voltage or momentary interruption occurs, the reactor has an unwanted trip by insertion of CARs and SORs even though the process systems are still in operation. HANARO was experienced in a nuisance trip as often as the unexpected voltage sag and/or momentary interruption occurs. We installed the voltage sag compensator on the power supply for CARs and SORs so as to prevent an unwanted trip. We undertook voltage sag assessment of the AC coil contactor which is a component of the power supply unit for the SORs. The compensation time is determined to be less than 1 sec in consideration of the reactor safety. This paper is concerned with the impact of the momentary interruption on the reactor and the effect of the voltage sag compensator

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.)

Atmospheric-pressure small-scale thermal-hydraulic experiment of a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, Kanji; Tamaki, Masayoshi; Imai, Satoshi; Kohketsu, Hideto; Anoda, Yoshinari; Murata, Hideo; Kukita, Yutaka.

1992-01-01

An experimental small-scale low-pressure setup of a PIUS (Process Inherent Ultimate Safety)-type reactor was used for the examination of the stability during normal operation such as startup and load following operation and of the safety during accidents such as loss-of-feedwater and pump runaway. Automatic feedback pump control system based on differential pressure at lower honeycomb density lock was quite effective to maintain the stratified interface between primary and pool water in the honeycomb density lock during normal operation. The process inherent ultimate safety characteristics of the PIUS-type reactor was confirmed with pump-trip scram at the pump speed limit for the various simulated accidents such as a loss-of-feedwater and pump runaway. (author)

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

Installation of the sag compensator for HANARO

International Nuclear Information System (INIS)

Kim, Hyung Kyoo; Jung, Hoan Sung; Lim, In Cheol; Ahn, Guk Hoon

2008-01-01

Electric power is essential for all industrial plants and also for nuclear facilities. HANARO is a research reactor which produces a 30MW thermal power. HANARO is designed to be tripped automatically when interruptions or some extent of sags occur. HANARO has the reactor regulation system(RRs) and reactor protection system(RPS). HANARO is designed so as to be tripped automatically by insertion of control absorber rods(CAR) and shut off rods(SOR). When voltage sag or momentary interruption occurs, the reactor has an unwanted trip by insertion of CARs and SORs even though the process systems are still in operation. HANARO was experienced in a nuisance trip as often as the unexpected voltage sag and/or momentary interruption occurs. We installed the voltage sag compensator voltage sag assessment of the AC coil contactor which is a component of the power supply unit for the SORs. The compensation time is determined to be less than 1 sec in consideration of the reactor safety. This paper is concerned with the impact of the momentary interruption on the reactor and the effect of the voltage sag compensator

Experimental Breeder Reactor-II automatic control-rod-drive system

International Nuclear Information System (INIS)

Christensen, L.J.

1983-01-01

A computer-controlled automatic control rod drive system (ACRDS) was designed and operated in EBR-II during reactor runs 121 and 122. The ACRDS was operated in a checkout mode during run 121 using a low worth control rod. During run 122 a high worth control rod was used to perform overpower transient tests as part of the LMFBR oxide fuels transient testing program. The testing program required an increase in power of 4 MW/s, a hold time of 12 minutes and a power decrease of 4 MW/s. During run 122, 13 power transients were performed

Wide-range nuclear reactor temperature control using automatically tuned fuzzy logic controller

International Nuclear Information System (INIS)

Ramaswamy, P.; Edwards, R.M.; Lee, K.Y.

1992-01-01

In this paper, a fuzzy logic controller design for optimal reactor temperature control is presented. Since fuzzy logic controllers rely on an expert's knowledge of the process, they are hard to optimize. An optimal controller is used in this paper as a reference model, and a Kalman filter is used to automatically determine the rules for the fuzzy logic controller. To demonstrate the robustness of this design, a nonlinear six-delayed-neutron-group plant is controlled using a fuzzy logic controller that utilizes estimated reactor temperatures from a one-delayed-neutron-group observer. The fuzzy logic controller displayed good stability and performance robustness characteristics for a wide range of operation

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

Automatic accounting of nuclear materials at WWER type reactor NPPs

International Nuclear Information System (INIS)

Babaev, N.S.; Poznyakov, N.L.; Strelkov, D.F.

1978-01-01

The possibilities of automatic accounting of nuclear materials at NPPs based on WWER reactors are considered. Organizational and technical principles of an automated system of accounting that takes into consideration IAEA requirements in conducting accounting documentation are proposed. A program is described for accounting materials using a BESM-6 computer. Operation of the program requires that all accounting data be recorded on conventional carriers of computer information (magnetic tapes, discs, perforated cards), which constitute the basic NPP accounting documents and may be directly used as initial data for a corresponding information program

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

An automatic device for sample insertion and extraction to/from reactor irradiation facilities

International Nuclear Information System (INIS)

Alloni, L.; Venturelli, A.; Meloni, S.

1990-01-01

At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

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)

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.)

Nuclear plant scram reduction

International Nuclear Information System (INIS)

Wiegle, H.R.

1986-01-01

The Nuclear Utility Management and Human Resources Committee (NUMARC) is a confederation of all 55 utilities with nuclear plants either in operation or under construction. NUMARC was formed in April 1984 by senior nuclear executives with hundreds of man-years of plant experience to improve (plant) performance and resolve NRC concerns. NUMARC has adopted 10 commitments in the areas of management, training, staffing and performance. One of these commitments is to strive to reduce automatic trips to 3 per year per unit for calendar year 1985 for plants in commercial operation greater than 3 years (with greater than 25% capacity factor). This goal applies to any unplanned automatic protection system trips at any time when the reactor is critical. Each utility has committed to develop methods to thoroughly evaluate all unplanned automatic trips to identify the root causes and formulate plans to correct the root causes thus reducing future unplanned scrams. As part of this program, the Institute of Nuclear Power Operations (INPO) collects and evaluates information on automatic reactor trips. It publishes the results of these evaluations to aid the industry to identify root causes and corrective actions

Emergency automatic commutation of the ventilation system of the RP-10 nuclear reactor

International Nuclear Information System (INIS)

Castillo, Walter; Corimanya, Mario; Ovalle, Edgar; Anaya, Olgger; Veramendi, Emilio

2013-01-01

The present paper summarizes the achievements in the design and implementation of a system for monitoring and automatic control of radioactive effluents from the chimney of the RP-10 reactor, using as hardware an Arduino UNO platform containing an ATMEGA 328 programmable micro controller to which has been added LCD screen to display the values, a keyboard and an EEPROM memory data, where the limit of the level of radiation is fixed. The radiation level in the air of the reactor hall, going up the chimney is counted by a radiation monitor called MAB1000, and data are supplied to the new system. When the radiation level is above the national and international standards, the new design makes work a relay, so that the ventilation system is automatically switched to operate in emergency condition, preventing the release of radioactive contaminants into the environment. After installing the new design, it was verified that removed by the radiation monitor MAB1000, value is identical to that shown in the new system. Additionally, the operation of the relay was tested successfully with radioactive sources to switch the ventilation system to the emergency condition. (authors).

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

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

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)

Automatic motion inhibit system for a nuclear power generating system

International Nuclear Information System (INIS)

Musick, C.R.; Torres, J.M.

1977-01-01

Disclosed is an automatic motion inhibit system for a nuclear power generating system for inhibiting automatic motion of the control elements to reduce reactor power in response to a turbine load reduction. The system generates a final reactor power level setpoint signal which is continuously compared with a reactor power signal. The final reactor power level setpoint is a setpoint within the capacity of the bypass valves to bypass steam which in no event is lower in value than the lower limit of automatic control of the reactor. If the final reactor power level setpoint is greater than the reactor power, an inhibit signal is generated to inhibit automatic control of the reactor. 6 claims, 5 figures

Effect of weather on pedestrian trip count and duration: City-scale evaluations using mobile phone application data.

Science.gov (United States)

Vanky, Anthony P; Verma, Santosh K; Courtney, Theodore K; Santi, Paolo; Ratti, Carlo

2017-12-01

We examined the association between meteorological (weather) conditions in a given locale and pedestrian trips frequency and duration, through the use of locative digital data. These associations were determined for seasonality, urban microclimate, and commuting. We analyzed GPS data from a broadly available activity tracking mobile phone application that automatically recorded 247,814 trips from 5432 unique users in Boston and 257,697 trips from 8256 users in San Francisco over a 50-week period. Generally, we observed increased air temperature and the presence of light cloud cover had a positive association with hourly trip frequency in both cities, regardless of seasonality. Temperature and weather conditions generally showed greater associations with weekend and discretionary travel, than with weekday and required travel. Weather conditions had minimal association with the duration of the trip, once the trip was initiated. The observed associations in some cases differed between the two cities. Our study illustrates the opportunity that emerging technology presents to study active transportation, and exposes new methods to wider consideration in preventive medicine.

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.)

Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

International Nuclear Information System (INIS)

Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I.

1999-12-01

One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (author)

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.

Indigenous technology development : seismic switch for nuclear reactors

International Nuclear Information System (INIS)

Varghese, Shiju; Shah, Jay; Limaye, P.K.; Soni, N.L; Patel, R.J.

2016-01-01

After Fukushima incident it has become a regulatory requirement to have automatic reactor trip on detection of earthquake beyond OBE level. Seismic Switches that meets the technical specifications required for nuclear reactor use were not available in the market. Hence, on Nuclear Power Corporation of India Ltd (NPCIL's) request, Refuelling Technology Division, BARC has developed Seismic Switches (electronic earthquake detectors) required for this application. Functionality of the system was successfully tested using a Shake Table. Two different designs of seismic switches have been developed. One is a microcontroller based system (digital) and the other is fully analogue electronics (analog) based. These switches are designed to meet the technical requirements of Class IA systems of nuclear reactors. It is also designed to meet other qualification tests such as EMI/EMC, climatic, vibration, and reliability requirements. In addition to nuclear industry seismic switches are having potential use in oil and gas, power plants, buildings and other industrial installations. These technologies are currently available for technology transfer and details are published in BARC website. This paper describes the requirements, principle of operation, and features and testing of the developed systems. (author)

Automatic recloser circuit breaker integrated with GSM technology for power system notification

Science.gov (United States)

Lada, M. Y.; Khiar, M. S. A.; Ghani, S. A.; Nawawi, M. R. M.; Rahim, N. H.; Sinar, L. O. M.

2015-05-01

Lightning is one type of transient faults that usually cause the circuit breaker in the distribution board trip due to overload current detection. The instant tripping condition in the circuit breakers clears the fault in the system. Unfortunately most circuit breakers system is manually operated. The power line will be effectively re-energized after the clearing fault process is finished. Auto-reclose circuit is used on the transmission line to carry out the duty of supplying quality electrical power to customers. In this project, an automatic reclose circuit breaker for low voltage usage is designed. The product description is the Auto Reclose Circuit Breaker (ARCB) will trip if the current sensor detects high current which exceeds the rated current for the miniature circuit breaker (MCB) used. Then the fault condition will be cleared automatically and return the power line to normal condition. The Global System for Mobile Communication (GSM) system will send SMS to the person in charge if the tripping occurs. If the over current occurs in three times, the system will fully trip (open circuit) and at the same time will send an SMS to the person in charge. In this project a 1 A is set as the rated current and any current exceeding a 1 A will cause the system to trip or interrupted. This system also provides an additional notification for user such as the emergency light and warning system.

Development of a protection system for research reactor based in Field Programmable Gate Array - FPGA; Desenvolvimento de sistema de protecao para reator nuclear de pesquisa baseado em Field Programmable Gate Array - FPGA

Energy Technology Data Exchange (ETDEWEB)

Martins, Roque Hudson da Silva

2016-07-01

This study presents a implementation purpose of a protection system for research nuclear reactors by using a programed device FPGA (Field Programmable Gate Array). As well as logic protection method involved on an automatic shutdown (TRIP) of a reactor, that ensure the security on such systems. These new control and operation mechanics are developed to guarantee that the security limits of a power plant are not exceeded, these mechanics can work isolated or in groups to safe guard the security levels. For this implementation to be completed, there will be presented the main aspects and concepts referred to protection systems, mostly about research nuclear reactors, with some applications terms exposed. The system proposed at this paper was developed following the VHDL (Very High Speed Integrated Circuits) hardware describing language, and the Modelsim software from Altera Software to program the automatic turning off routines, and hypothetical simulations for such. The results show that for every software application for supporting nuclear reactors, like security devices, they have to meet the IEC 60880 criteria. This paper have great importance, seeing that nuclear reactor security systems, are a basic element for ensure the reactor security. (author)

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)

Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

1999-12-01

One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

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…

Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR

International Nuclear Information System (INIS)

Kim, M. W.; Kim, C. H.; Hong, I. S.

2005-02-01

The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse

Effect of weather on pedestrian trip count and duration: City-scale evaluations using mobile phone application data

Directory of Open Access Journals (Sweden)

Anthony P. Vanky

2017-12-01

Full Text Available We examined the association between meteorological (weather conditions in a given locale and pedestrian trips frequency and duration, through the use of locative digital data. These associations were determined for seasonality, urban microclimate, and commuting. We analyzed GPS data from a broadly available activity tracking mobile phone application that automatically recorded 247,814 trips from 5432 unique users in Boston and 257,697 trips from 8256 users in San Francisco over a 50-week period. Generally, we observed increased air temperature and the presence of light cloud cover had a positive association with hourly trip frequency in both cities, regardless of seasonality. Temperature and weather conditions generally showed greater associations with weekend and discretionary travel, than with weekday and required travel. Weather conditions had minimal association with the duration of the trip, once the trip was initiated. The observed associations in some cases differed between the two cities. Our study illustrates the opportunity that emerging technology presents to study active transportation, and exposes new methods to wider consideration in preventive medicine. Keywords: Weather, Pedestrian activity, Walking, Weather conditions and active transportation, Microclimates, Spatial behavior, Mobile phones, Locative data, Emerging technology, Big data

Automatic control system at the ''Loviisa'' NPP

International Nuclear Information System (INIS)

Kukhtevich, I.V.; Mal'tsev, B.K.; Sergievskaya, E.N.

1980-01-01

Automatic control system of the Loviisa-1 NPP (Finland) is described. According to operation conditions of Finland power system the Loviisa-1 NPP must operate in the mode of week and day control of loading schedule and participate in current control of power system frequency and capacity. With provision for these requirements NPP is equipped with the all-regime system for automatic control functioning during reactor start-up, shut-down, in normal and transient regimes and in emergency situations. The automatic control system includes: a data subsystem, an automatic control subsystem, a discrete control subsystem including remote, a subsystem for reactor control and protection and overall station system of protections: control and dosimetry inside the reactor. Structures of a data-computer complex, discrete control subsystems, reactor control and protection systems, neutron flux control system, inside-reactor control system, station protection system and system for control of fuel element tightness are presented in short. Two-year experience of the NPP operation confirmed advisability of the chosen volume of automatization. The Loviisa-1 NPP operates successfully in the mode of the week and day control of supervisor schedule and current control of frequency (short-term control)

Automatic control of nuclear power plants

International Nuclear Information System (INIS)

Jover, P.

1976-01-01

The fundamental concepts in automatic control are surveyed, and the purpose of the automatic control of pressurized water reactors is given. The response characteristics for the main components are then studied and block diagrams are given for the main control loops (turbine, steam generator, and nuclear reactors) [fr

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

Automatic control of scale range applied for analog study of reactor kinetics

International Nuclear Information System (INIS)

Sergent, O.; Tellier, N.

1967-01-01

We study the response of a reactor, initially in a sub-critical state, for linear release of reactivity obeying to the following criteria, a rod drop comes in 10 seconds after the moment when the neutron power becomess equal to 10 -3 times the nominal power. We are interested in the maximum reactivity reached and in the energy released during the power excursion. For the power varying in a range from 1 to 10 10 we have used the method of automatic change scale which was installed and described in a previous report [fr

An automatic tuning method of a fuzzy logic controller for nuclear reactors

International Nuclear Information System (INIS)

Ramaswamy, P.; Lee, K.Y.; Edwards, R.M.

1993-01-01

The design and evaluation by simulation of an automatically tuned fuzzy logic controller is presented. Typically, fuzzy logic controllers are designed based on an expert's knowledge of the process. However, this approach has its limitations in the fact that the controller is hard to optimize or tune to get the desired control action. A method to automate the tuning process using a simplified Kalman filter approach is presented for the fuzzy logic controller to track a suitable reference trajectory. Here, for purposes of illustration an optimal controller's response is used as a reference trajectory to determine automatically the rules for the fuzzy logic controller. To demonstrate the robustness of this design approach, a nonlinear six-delayed neutron group plant is controlled using a fuzzy logic controller that utilizes estimated reactor temperatures from a one-delayed neutron group observer. The fuzzy logic controller displayed good stability and performance robustness characteristics for a wide range of operation

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)

Development of advanced automatic operation system for nuclear ship. 1. Perfect automatic normal operation

International Nuclear Information System (INIS)

Nakazawa, Toshio; Yabuuti, Noriaki; Takahashi, Hiroki; Shimazaki, Junya

1999-02-01

Development of operation support system such as automatic operating system and anomaly diagnosis systems of nuclear reactor is very important in practical nuclear ship because of a limited number of operators and severe conditions in which receiving support from others in a case of accident is very difficult. The goal of development of the operation support systems is to realize the perfect automatic control system in a series of normal operation from the reactor start-up to the shutdown. The automatic control system for the normal operation has been developed based on operating experiences of the first Japanese nuclear ship 'Mutsu'. Automation technique was verified by 'Mutsu' plant data at manual operation. Fully automatic control of start-up and shutdown operations was achieved by setting the desired value of operation and the limiting value of parameter fluctuation, and by making the operation program of the principal equipment such as the main coolant pump and the heaters. This report presents the automatic operation system developed for the start-up and the shutdown of reactor and the verification of the system using the Nuclear Ship Engineering Simulator System. (author)

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)

Design and development of indigenous seismic switch for nuclear reactors

International Nuclear Information System (INIS)

Varghese, Shiju; Shah, Jay; Limaye, P.K.; Soni, N.L; Patel, R.J.

2016-01-01

After Fukushima incident it has become a regulatory requirement to have automatic reactor trip on detection of earthquake beyond OBE level. Seismic Switches that meets the technical specifications required for nuclear reactor use were not available in the market. Hence, on Nuclear Power Corporation of India Ltd (NPCIL's) request, Refuelling Technology Division, BARC has developed Seismic Switches (electronic earthquake detectors) required for this application. Functionality of the system was successfully tested using a Shake Table. Two different designs of seismic switches have been developed. One is a microcontroller based system (digital) and the other is fully analogue electronics (analog) based. These switches are designed to meet the technical requirements of Class IA systems of nuclear reactors. It is also designed to meet other qualification tests such as EMI/EMC, climatic, vibration, and reliability requirements. In addition to nuclear industry seismic switches are having potential use in oil and gas, power plants, buildings and other industrial installations. These technologies are currently available for technology transfer and details are published in BARC website. This paper describes the requirements, principle of operation and features and testing of the developed systems. (author)

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

Automatic systems for opening and closing reactor vessels, steam generators, and pressurizers

International Nuclear Information System (INIS)

Samblat, C.

1990-01-01

The need for shorter working assignments, reduced dose rates and less time consumption have caused Electricite de France and Framatome to automate the entire procedure of opening and closing the main components in the primary system, such as the reactor vessel, steam generator, and pressurizer. The experience accumulated by the two companies in more than 300 annual revisions of nuclear generating units worldwide has been used as a basis for automating all bolt opening and closing steps as well as cleaning processes. The machines and automatic systems currently in operation are the result of extensive studies and practical tests. (orig.) [de

Availability verification of information for human system interface in automatic SG level control using activity diagram

Energy Technology Data Exchange (ETDEWEB)

Nuraslinda, Anuar; Kim, Dong Young; Kim, Jong Hyun [KEPCO International Nuclear Graduate School, Uljugun (Korea, Republic of)

2012-10-15

Steam Generator (SG) level control system in OPR 1000 is one of representative automatic systems that falls under the Supervisory Control level in Endsley's taxonomy. Supervisory control of automated systems is classified as a form of out of the loop (OOTL) performance due to passive involvement in the systems operation, which could lead to loss of situation awareness (SA). There was a reported event, which was caused by inadequate human automation communication that contributed to an unexpected reactor trip in July 2005. A high SG level trip occurred in Yeonggwang (YGN) Unit 6 Nuclear Power Plant (NPP) due to human operator failure to recognize the need to change the control mode of the economizer valve controller (EVC) to manual mode during swap over (the transition from low power mode to high power mode) after the loss of offsite power (LOOP) event was recovered. This paper models the human system interaction in NPP SG level control system using Unified Modeling Language (UML) Activity Diagram. Then, it identifies the missing information for operators in the OPR1000 Main Control Room (MCR) and suggests some means of improving the human system interaction.

Availability verification of information for human system interface in automatic SG level control using activity diagram

International Nuclear Information System (INIS)

Nuraslinda, Anuar; Kim, Dong Young; Kim, Jong Hyun

2012-01-01

Steam Generator (SG) level control system in OPR 1000 is one of representative automatic systems that falls under the Supervisory Control level in Endsley's taxonomy. Supervisory control of automated systems is classified as a form of out of the loop (OOTL) performance due to passive involvement in the systems operation, which could lead to loss of situation awareness (SA). There was a reported event, which was caused by inadequate human automation communication that contributed to an unexpected reactor trip in July 2005. A high SG level trip occurred in Yeonggwang (YGN) Unit 6 Nuclear Power Plant (NPP) due to human operator failure to recognize the need to change the control mode of the economizer valve controller (EVC) to manual mode during swap over (the transition from low power mode to high power mode) after the loss of offsite power (LOOP) event was recovered. This paper models the human system interaction in NPP SG level control system using Unified Modeling Language (UML) Activity Diagram. Then, it identifies the missing information for operators in the OPR1000 Main Control Room (MCR) and suggests some means of improving the human system interaction

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

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)

Development and implementation of an automatic control algorithm for the University of Utah nuclear reactor

International Nuclear Information System (INIS)

Crawford, Kevan C.; Sandquist, Gary M.

1990-01-01

The emphasis of this work is the development and implementation of an automatic control philosophy which uses the classical operational philosophies as a foundation. Three control algorithms were derived based on various simplifying assumptions. Two of the algorithms were tested in computer simulations. After realizing the insensitivity of the system to the simplifications, the most reduced form of the algorithms was implemented on the computer control system at the University of Utah (UNEL). Since the operational philosophies have a higher priority than automatic control, they determine when automatic control may be utilized. Unlike the operational philosophies, automatic control is not concerned with component failures. The object of this philosophy is the movement of absorber rods to produce a requested power. When the current power level is compared to the requested power level, an error may be detected which will require the movement of a control rod to correct the error. The automatic control philosophy adds another dimension to the classical operational philosophies. Using this philosophy, normal operator interactions with the computer would be limited only to run parameters such as power, period, and run time. This eliminates subjective judgements, objective judgements under pressure, and distractions to the operator and insures the reactor will be operated in a safe and controlled manner as well as providing reproducible operations

Automatic plasma control in magnetic traps

International Nuclear Information System (INIS)

Samojlenko, Y.; Chuyanov, V.

1984-01-01

Hot plasma is essentially in thermodynamic non-steady state. Automatic plasma control basically means monitoring deviations from steady state and producing a suitable magnetic or electric field which brings the plasma back to its original state. Briefly described are two systems of automatic plasma control: control with a magnetic field using a negative impedance circuit, and control using an electric field. It appears that systems of automatic plasma stabilization will be an indispensable component of the fusion reactor and its possibilities will in many ways determine the reactor economy. (Ha)

Theoretical study of an on-off automatic control system for a nuclear reactor

International Nuclear Information System (INIS)

Menezes, J.; Jover, P.

1964-01-01

The automatic control system designed for a high flux nuclear reactor is of the 'constant speed-dead zone' type, in which the control rod is run at normal speed in the required direction when the error signal overrides a preset level. The study of the closed loop absolute stability was carried out with the describing function method. An analog computer study yielded the optimal values of the setting parameters (relay hysteresis, motor response time), which lead to a minimization of the control steps frequency when the reactivity varies slowly with the xenon poisoning. (authors) [fr

Automatic operation device for control rods

International Nuclear Information System (INIS)

Sekimizu, Koichi

1984-01-01

Purpose: To enable automatic operation of control rods based on the reactor operation planning, and particularly, to decrease the operator's load upon start up and shutdown of the reactor. Constitution: Operation plannings, demand for the automatic operation, break point setting value, power and reactor core flow rate change, demand for operation interrupt, demand for restart, demand for forecasting and the like are inputted to an input device, and an overall judging device performs a long-term forecast as far as the break point by a long-term forecasting device based on the operation plannings. The automatic reactor operation or the like is carried out based on the long-term forecasting and the short time forecasting is performed by the change in the reactor core status due to the control rod operation sequence based on the control rod pattern and the operation planning. Then, it is judged if the operation for the intended control rod is possible or not based on the result of the short time forecasting. (Aizawa, K.)

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.

Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

International Nuclear Information System (INIS)

Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

1979-01-01

The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

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.)

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)

Detection of a regulating valve closure failure during review of recorded data after an automatic reactor shut down. Incident at the NPP Beznau-1, 27 April 1995

International Nuclear Information System (INIS)

Deutschmann, H.

1996-01-01

After recognizing a leak in the oil system of the running main feedwater pump 1 during rated power operation of the plant the operator changed feedwater supply manually to the stand-by pump 2. A short time later pump 2 was automatically tripped by the signal ''low oil pressure''. Immediate reduction of the reactor power by the operator was not successful because the scram signal ''low steam generator level and mismatch of steam/feedwater flow'' occurred and scram was actuated. In this plant a special operating feature, actuated by the scram signal, is implemented to reduce steam release to atmosphere in case of scram. The signal ''scram and average primary Temperature >287 deg. C opens the feedwater regulating valves, and later, if the average primary temperature decreases to <287 deg. C, they reclose by a redundant signal. In the experienced event, after the scram actuation, in the steam generator A a feedwater overfill occurred. The overfill protection tripped the operating feedwater pumps (main feedwater pump 3 and two auxiliary feedwater pumps). The large injection of water produced an overcooling of the primary with isolation of the volume control system outlet of the primary. The operator repaired the defective oil coolers of the feedwater pumps and restarted the plant. At that time, he had not recognized, that the plant response, which caused the steam generator overfill, was wrong. One day later, as all the recorded data were reviewed in more detail, it was found that the closure time of the feedwater regulating valve to steam generator A was much longer than designed (19 s instead 7 s). The operator requested an LCO for continued operation in spite of the fact, that the closure time was not fixed in the Technical specification. 3 figs

Steam leak detection in advance reactors via acoustics method

International Nuclear Information System (INIS)

Singh, Raj Kumar; Rao, A. Rama

2011-01-01

Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T.

1997-01-01

The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

Reactor core control device

International Nuclear Information System (INIS)

Sano, Hiroki

1998-01-01

The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

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

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.

Application of an automatic pattern recognition for aleatory signals for the surveillance of nuclear reactor and rotating machinery

International Nuclear Information System (INIS)

Nascimento, J.A. do.

1982-02-01

An automatic pattern recognition program PSDREC, developed for the surveillance of nuclear reactor and rotating machinery is described and the relevant theory is outlined. Pattern recognition analysis of noise signals is a powerful technique for assessing 'system normality' in dynamic systems. This program, with applies 8 statistical tests to calculated power spectral density (PSD) distribution, was earlier installed in a PDP-11/45 computer at IPEN. To analyse recorded signals from three systems, namely an operational BWR power reactor (neutron signals), a water pump and a diesel engine (vibration signals) this technique was used. Results of the tests are considered satisfactory. (Author) [pt

Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

International Nuclear Information System (INIS)

Ramirez G, C.; Chavez M, C.

2012-10-01

The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

Development of a simple method for classifying the degree of importance of components in nuclear power plants using probabilistic analysis technique

International Nuclear Information System (INIS)

Shimada, Yoshio; Miyazaki, Takamasa

2006-01-01

In order to analyze large amounts of trouble information of overseas nuclear power plants, it is necessary to select information that is significant in terms of both safety and reliability. In this research, a method of efficiently and simply classifying degrees of importance of components in terms of safety and reliability while paying attention to root-cause components appearing in the information was developed. Regarding safety, the reactor core damage frequency (CDF), which is used in the probabilistic analysis of a reactor, was used. Regarding reliability, the automatic plant trip probability (APTP), which is used in the probabilistic analysis of automatic reactor trips, was used. These two aspects were reflected in the development of criteria for classifying degrees of importance of components. By applying these criteria, a method of quantitatively and simply judging the significance of trouble information of overseas nuclear power plants was developed. (author)

Method of extracting significant trouble information of nuclear power plants using probabilistic analysis technique

International Nuclear Information System (INIS)

Shimada, Yoshio; Miyazaki, Takamasa

2005-01-01

In order to analyze and evaluate large amounts of trouble information of overseas nuclear power plants, it is necessary to select information that is significant in terms of both safety and reliability. In this research, a method of efficiently and simply classifying degrees of importance of components in terms of safety and reliability while paying attention to root-cause components appearing in the information was developed. Regarding safety, the reactor core damage frequency (CDF), which is used in the probabilistic analysis of a reactor, was used. Regarding reliability, the automatic plant trip probability (APTP), which is used in the probabilistic analysis of automatic reactor trips, was used. These two aspects were reflected in the development of criteria for classifying degrees of importance of components. By applying these criteria, a simple method of extracting significant trouble information of overseas nuclear power plants was developed. (author)

LARA. Localization of an automatized refueling machine by acoustical sounding in breeder reactors - implementation of artificial intelligence techniques

International Nuclear Information System (INIS)

Lhuillier, C.; Malvache, P.

1987-01-01

The automatic control of the machine which handles the nuclear subassemblies in fast neutron reactors requires autonomous perception and decision tools. An acoustical device allows the machine to position in the work area. Artificial intelligence techniques are implemented to interpret the data: pattern recognition, scene analysis. The localization process is managed by an expert system. 6 refs.; 8 figs

Reactor power control device

International Nuclear Information System (INIS)

Doi, Kazuyori.

1981-01-01

Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

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.

Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

International Nuclear Information System (INIS)

Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

2012-01-01

The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

Analysis of SBO ATWS for Maanshan PWR

Energy Technology Data Exchange (ETDEWEB)

Chen, Che-Hao; Chen, Shao-Wen [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Nuclear and New Energy Education and Research Foundation, Hsinchu, Taiwan (China); Lin, Hao-Tzu [Atomic Energy Council, Taoyuan, Taiwan (China). Inst. of Nuclear Energy Research

2015-11-15

Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

Emergency automatic commutation of the ventilation system of the RP-10 nuclear reactor; Conmutacion automatica de emergencia del sistema de ventilacion del reactor nuclear RP-10

Energy Technology Data Exchange (ETDEWEB)

Castillo, Walter; Corimanya, Mario; Ovalle, Edgar; Anaya, Olgger; Veramendi, Emilio [Direccion de Produccion, Instituto Peruano de Energia Nuclear, Av. Jose Saco km 12.5, Carabayllo, Lima (Peru)

2013-07-01

The present paper summarizes the achievements in the design and implementation of a system for monitoring and automatic control of radioactive effluents from the chimney of the RP-10 reactor, using as hardware an Arduino UNO platform containing an ATMEGA 328 programmable micro controller to which has been added LCD screen to display the values, a keyboard and an EEPROM memory data, where the limit of the level of radiation is fixed. The radiation level in the air of the reactor hall, going up the chimney is counted by a radiation monitor called MAB1000, and data are supplied to the new system. When the radiation level is above the national and international standards, the new design makes work a relay, so that the ventilation system is automatically switched to operate in emergency condition, preventing the release of radioactive contaminants into the environment. After installing the new design, it was verified that removed by the radiation monitor MAB1000, value is identical to that shown in the new system. Additionally, the operation of the relay was tested successfully with radioactive sources to switch the ventilation system to the emergency condition. (authors).

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

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...

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

Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

International Nuclear Information System (INIS)

Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

1999-10-01

The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

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

Investigation of the Bilibin reactor operation in the regime of automatic power and frequency control in isolated power system

International Nuclear Information System (INIS)

Sankovskij, G.A.; Molochkov, V.I.; Dolgov, V.V.; Soldatov, G.E.; Minashin, M.E.

1981-01-01

The results of experimental investigations of the power unit operation of the Bilibin nuclear power and heating plant (BNPHP) in the regime of automatic power and frequency control in an isolated power system are presented. The BNPHP comprises four similar power units. Each unit includes a steam generating setup - the channel water-graphite reactor with tubular fuel elements with natural circulation of boiling water at all the power levels as well as a turbosetup with two heat selectors and a turbogenerator. The turbine operates on dry saturated steam (with intermediate separation) which is brought from the drum-separator of the reactor natural circulation circuit. The BNPHP operates according to the controller schedule since the start-up of the first power unit. The BNPHP unit power varifies within the 50-100% range 3-4 times per day (by the number of maxima in the schedule of the power system loadings). Two design flowsheets of the unit power control and dynamic characteristics of the system for both vatiants are considered. It is concluded that both investigated automatic control systems are seviceable and deviations of the reactor parameters within the transients are not dangerous for heat release from the core. The plant is better shielded from external mainly short-term perturbations coming from the power system when the system operates in accordance with the first variant of the flowsheet [ru

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.)

Control of WWER-440 nuclear reactor

International Nuclear Information System (INIS)

Wagner, K.; Drab, F.; Grof, V.

1978-01-01

The V-1 reactor control systems are described. The data acquisition and processing system fulfils four main functions, ie., reactor start-up and power increase to 10% of the rated power, automatic power control within 3% and 110% of the rated power, reactivity compensation, and reactor protection. The automatic control system ensures constant steam pressure maintained with an accuracy of +-0.05 MPa. Reactivity compensation and spatial power distribution is mainly safeguarded by boric acid control. The V-1 reactor protection system has four levels of accident protection depending on the gravity of the failure. The philosophy of automation of the V-1 reactor control and protection system is based on autonomous automatic controlers and on the direct control of the individual sets and technological equipment. In conclusion, development trends are briefly outlined of control and protection systems of light water reactor power plants. (Z.M.)

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

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)

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

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

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)

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

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

A completely automatic operation type super-safe fast reactor, RAPID. Its application to dispersion source on lunar and earth surfaces

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Kawasaki, Akira; Iwamura, Takamichi

2002-01-01

At a viewpoint of flexible measures to future electric power demands, expectation onto a small-scale reactor for dispersion source is increasing gradually. This is thought to increase its importance not only for a source at proximity of its market in advanced nations but also for the one in developing nations. A study on development of the completely automatic operation type super-safe fast reactor, RAPID (refueling by all pins integrated design) has been carried out as a part of the nuclear energy basic research promoting system under three years project since 1999 by a trust of the Japan Atomic Energy Research Institute to a group of the Central Research Institute of Electric Power Industry (CRIEPI) and so on. As the reactor is a lithium cooled fast reactor with 200 Kw of electric output supposing to use at lunar surface, it can be applied to a super-small scale nuclear reactor on the earth, and has feasibility to become a new option of future nuclear power generation. On the other hand, CRIEPI has investigated on various types of fast reactors (RAPID series) for fast reactor for dispersion source on the earth. Here was introduced on such super-safe fast reactors at a center of RAPID-L. (G.K.)

IE Information Notice No. 85-75: Improperly installed instrumentation, inadequate quality control and inadequate postmodification testing

International Nuclear Information System (INIS)

Jordan, E.L.

1992-01-01

On June 10, 1985, the licensee informed the NRC Resident Inspector that for approximately 5 days LaSalle Unit 2 had been without the capability of automatic actuation of emergency core cooling (ECCS) and that for approximately 3 days during this period the plant had been without secondary containment integrity. The major cause of this condition was improper installation (the variable and reference legs were reversed) of the two reactor vessel level actuation switches which control Division 1 automatic depressurization system (ADS), low pressure core spray (LPCS), and reactor core isolation cooling (RCIC). On July 20, 1985, the Trojan Nuclear Power Plant tripped from 100% power because of a turbine trip that was caused by the loss of the unit auxiliary transformer. All systems functioned normally except that low suction pressure caused one auxiliary feedwater pump to trip and then the other auxiliary feedwater pump to trip after restart of the first auxiliary feedwater pump. The cause of the trips of the auxiliary feedwater pumps can be traced back to improper postmodification adjustment and inadequate postmodification testing following retrofit of environmentally qualified controllers for the auxiliary feedwater system. The auxiliary feedwater pump trips on low suction pressure were caused by excessive combined flow from the two auxiliary feedwater pumps that draw from a single header from the condensate storage tank. The flow control valves were open farther than required after new environmentally qualified controllers had been installed during a recent refueling outage

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...

ATIPS: Automatic Travel Itinerary Planning System for Domestic Areas

Science.gov (United States)

2016-01-01

Leisure travel has become a topic of great interest to Taiwanese residents in recent years. Most residents expect to be able to relax on a vacation during the holidays; however, the complicated procedure of travel itinerary planning is often discouraging and leads them to abandon the idea of traveling. In this paper, we design an automatic travel itinerary planning system for the domestic area (ATIPS) using an algorithm to automatically plan a domestic travel itinerary based on user intentions that allows users to minimize the process of trip planning. Simply by entering the travel time, the departure point, and the destination location, the system can automatically generate a travel itinerary. According to the results of the experiments, 70% of users were satisfied with the result of our system, and 82% of users were satisfied with the automatic user preference learning mechanism of ATIPS. Our algorithm also provides a framework for substituting modules or weights and offers a new method for travel planning. PMID:26839529

ATIPS: Automatic Travel Itinerary Planning System for Domestic Areas

Directory of Open Access Journals (Sweden)

Hsien-Tsung Chang

2016-01-01

Full Text Available Leisure travel has become a topic of great interest to Taiwanese residents in recent years. Most residents expect to be able to relax on a vacation during the holidays; however, the complicated procedure of travel itinerary planning is often discouraging and leads them to abandon the idea of traveling. In this paper, we design an automatic travel itinerary planning system for the domestic area (ATIPS using an algorithm to automatically plan a domestic travel itinerary based on user intentions that allows users to minimize the process of trip planning. Simply by entering the travel time, the departure point, and the destination location, the system can automatically generate a travel itinerary. According to the results of the experiments, 70% of users were satisfied with the result of our system, and 82% of users were satisfied with the automatic user preference learning mechanism of ATIPS. Our algorithm also provides a framework for substituting modules or weights and offers a new method for travel planning.

ATIPS: Automatic Travel Itinerary Planning System for Domestic Areas.

Science.gov (United States)

Chang, Hsien-Tsung; Chang, Yi-Ming; Tsai, Meng-Tze

2016-01-01

Leisure travel has become a topic of great interest to Taiwanese residents in recent years. Most residents expect to be able to relax on a vacation during the holidays; however, the complicated procedure of travel itinerary planning is often discouraging and leads them to abandon the idea of traveling. In this paper, we design an automatic travel itinerary planning system for the domestic area (ATIPS) using an algorithm to automatically plan a domestic travel itinerary based on user intentions that allows users to minimize the process of trip planning. Simply by entering the travel time, the departure point, and the destination location, the system can automatically generate a travel itinerary. According to the results of the experiments, 70% of users were satisfied with the result of our system, and 82% of users were satisfied with the automatic user preference learning mechanism of ATIPS. Our algorithm also provides a framework for substituting modules or weights and offers a new method for travel planning.

Automatic loading pattern optimization tool for Loviisa VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Kuopanportti, Jaakko [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

2013-09-15

An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies' operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution. (orig.)

Automatic loading pattern optimization tool for Loviisa VVER-440 reactors

International Nuclear Information System (INIS)

Kuopanportti, Jaakko

2013-01-01

An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies' operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution. (orig.)

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)

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.

Information Flow Analysis for Human-System Interaction in the SG Level Control

International Nuclear Information System (INIS)

Kim, Jong Hyun; Shin, Yeong Cheol

2008-01-01

Interaction between automatic control and operators is one of main issues in the application of automation technology. Inappropriate information from automatic control systems causes unexpected problems in human-automation collaboration. Poor information becomes critical, especially when the operator takes over the control from an automation system. Operators cannot properly handle the situation transferred from the automatic mode because of inadequate situation awareness, if the operator is out-of-the loop and the automatic control system fails. Some cases of unplanned reactor trips during the transition between the manual mode and the automatic mode are reported in nuclear power plants (NPPs). Among unplanned reactor trips since 2002, two cases were partially caused by automation-related failures of steam generator (SG) level control. This paper conducts information flow analysis to identify information and control requirement for human-system interaction of SG level control. At first, this paper identifies the level of automation in SG level control systems and then function allocation between system control and human operators. Then information flow analysis for monitoring and transition of automation is performed by adapting job process chart. Information and control requirements will be useful as an input for the human-system interface (HSI) design of SG level control

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.

Control device for start-up of reactor depressurization system

International Nuclear Information System (INIS)

Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

1998-01-01

The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

SPV Analysis of CEDMCS in Advanced Power Reactors

Energy Technology Data Exchange (ETDEWEB)

Awwal, Arigi M.; Emmanuel, Efenji A. Emmanuel; Faragalla, Mohamed M.; Lee, Yong-kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

Single Point Vulnerability (SPV) is a component whose failure would directly cause an automatic or manual reactor scram or turbine trip. Although some power plants do not consider the cause of any reduction in power as SPV, others consider components that cause a reduction in power of as low as 2% as SPV. The Control Element Drive Mechanism Control System (CEDMCS) controls and regulates power supplied to drive the control rods with the Control Element Drive Mechanism (CEDM). A 4-coil CEDM is used in the newly built Advanced Power Reactor (APR) 1400 plant, while a new CEDMCS for 3-coil CEDM has been designed to be deployed to another APR1400 plant. This paper shows an approach to evaluate the SPVs that may be available in either of these two systems. System A design has employed a fail-safe concept to its design with less redundancies while System B design provides redundancy and design change although this comes at a high price for the Utility. The System B design has improved reliability but not necessarily eliminating the SPV items. Naturally, the cost of a new redundant system will be more. However, future work will examine the economic effect of the new system considering the operating experiences of power plants on the CEDMCS (i.e. SCRAM rates and power outage cost)

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.

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.

The computerized reactor period measurement system for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1996-01-01

The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

Experience with high percent step load decrease from full power in NPP Krsko

International Nuclear Information System (INIS)

Vukovic, V.

2000-01-01

The control system of NPP Kriko, is designed to automatically control the reactor in the power range between 15 and 100 percent of rated power for the following designed transients; - 10 percent step change in load; 5 percent per minute loading and unloading; step full load decrease with the aid of automatically initiated and controlled steam dump. Because station operation below 15 percent of rated power is designed for a period of time during startup or standby conditions, automatic control below 15 percent is not provided. The steam dump accomplishes the following functional tasks: it permits the nuclear plants to accept a sudden 95 percent loss of load without incurring reactor trip; it removes stored energy and residual heat following a reactor trip and brings the plant to equilibrium no-load conditions without actuation of the steam generator safety valves; it permits control of the steam generator pressure at no-load conditions and permits a manually controlled cooldown of the plant. The first two functional tasks are controlled by Tavg. The third is controlled by steam pressure. Interlocks minimise any possibility of an inadvertent actuation of steam dump system. This paper discusses relationships between designed (described) characteristics of plant and the data which are obtained during startup and/or first ten years of operation. (author)

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.)

AUTOLOAD, an automatic optimal pressurized water reactor reload design system with an expert module

International Nuclear Information System (INIS)

Li, Z.; Levine, S.H.

1994-01-01

An automatic optimal pressurized water reactor (PWR) reload design expert system AUTOLOAD has been developed. It employs two important new techniques. The first is a new loading priority scheme that defines the optimal placement of the fuel in the core that has the maximum end-of-cycle state k eff . The second is a new power-shape-driven progressive iteration method for automatically determining the burnable poison (BP) loading in the fresh fuel assemblies. The Haling power distribution is used in converting the theoretically optimal solution into the practical design, which meets the design constraints for the given fuel assemblies. AUTOLOAD is a combination of C and FORTRAN languages. It requires only the required cycle length, the maximum peak normalized power, the BP type, the number of fresh fuel assemblies, the assembly burnup, and BP histories of the available fuel assemblies as its input. Knowledge-based modules have been built into the expert system computer code to perform all of the tasks involved in reloading a PWR. AUTOLOAD takes only ∼ 30 CPU min on an IBM 3090 600s mainframe to accomplish a practical reload design. A maximum of 12.5% fresh fuel enrichment saving is observed compared with the core used by the utility

Automatic boiling water reactor control rod pattern design using particle swarm optimization algorithm and local search

Energy Technology Data Exchange (ETDEWEB)

Wang, Cheng-Der, E-mail: jdwang@iner.gov.tw [Nuclear Engineering Division, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Chaung [National Tsing Hua University, Department of Engineering and System Science, 101, Section 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China)

2013-02-15

Highlights: ► The PSO algorithm was adopted to automatically design a BWR CRP. ► The local search procedure was added to improve the result of PSO algorithm. ► The results show that the obtained CRP is the same good as that in the previous work. -- Abstract: This study developed a method for the automatic design of a boiling water reactor (BWR) control rod pattern (CRP) using the particle swarm optimization (PSO) algorithm. The PSO algorithm is more random compared to the rank-based ant system (RAS) that was used to solve the same BWR CRP design problem in the previous work. In addition, the local search procedure was used to make improvements after PSO, by adding the single control rod (CR) effect. The design goal was to obtain the CRP so that the thermal limits and shutdown margin would satisfy the design requirement and the cycle length, which is implicitly controlled by the axial power distribution, would be acceptable. The results showed that the same acceptable CRP found in the previous work could be obtained.

Electricity generation by nuclear fission reactor and closed cycle gas turbines, with core automatically shut down by coolant flow failure and dropped out of plant for sealing if temperature is excessive

International Nuclear Information System (INIS)

Pedrick, A.P.

1976-01-01

A reactor system is described in which if there is a failure of coolant flow the core automatically drops down to its control rods, so that criticality is reduced, but if the temperature of the core still stays dangerously high the core is allowed to drop down a deep shaft. Concrete blocks automatically come together after the ejected reactor core has moved past them to prevent the escape of radiation or radioactive material, until such time that the core temperature has dropped to a level that it can, with safety, be returned to its normal position in the plant. (U.K.)

Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

International Nuclear Information System (INIS)

Takeda, Takeshi; Furusawa, Takayuki

2001-07-01

Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

RIMACS, Reactor Inspection Main Control System

International Nuclear Information System (INIS)

2008-01-01

1 - Description of program or function: RIMACS prepares for automatic inspection files on each inspection item for the reactor. These automatic inspection files provide the data to move RIROB (Reactor Inspection Robot) with laser by interpreting the coordinates of LASPO (Laser Positioner) and the laser detecting device of RIROB in three dimensional space. In addition, when RIROB arrives at the inspecting location, the files provide all values of the manipulator's motions to acquire the ultrasonic data. RIMACS provides various modules in order to perform these complex functions, and the functions are programmed on graphic user interface for the convenience of the user. RIMACS provides various functions, such as insertion of reactor production data, selection of the reactor for inspection, the creation of automatic inspection file, the selection of the inspection item, inspection simulation, and automatic inspection procedures. It also provides all other functions, which are necessary for the inspection, such as operating program download and manual control of LASPO and RIROB, the inspection simulation and the inspection status display by means of the graphic screen, and SODAS (ultra-Sonic Data Acquisition System) drive verification. 2 - Methods: Moving path and operation procedures for inspection robot are generated automatically with Kinematics algorithm. 3 - Restrictions on the complexity of the problem: A graphics display with MS-Window capability is required

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

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...

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

Knowledge-based full-automatic control system for a nuclear ship reactor

International Nuclear Information System (INIS)

Shimazaki, J.; Nakazawa, T.; Yabuuchi, N.

2000-01-01

Plant operations aboard nuclear ships require quick judgements and actions due to changing marine conditions such as wind, waves and currents. Furthermore, additional human support is not available for nuclear ship operation at sea, so advanced automatic operations are necessary to reduce the number of operators required finally. Therefore, an advanced automatic operating system has been developed based on operational knowledge of nuclear ship 'Mutsu' plant. The advanced automatic operating system includes both the automatic operation system and the operator-support system which assists operators in completing actions during plant accidents, anomaly diagnosis and plant supervision. These system are largely being developed using artificial intelligent techniques such as neural network, fuzzy logic and knowledge-based expert. The automatic operation system is fundamentally based upon application of an operator's knowledge of both normal (start-up to rated power level) and abnormal (after scram) operations. Comparing plant behaviors from start-up to power level by the automatic operation with by 'Mutsu' manual operation, stable automatic operation was obtained almost same as manual operation within all operating limits. The abnormal automatic system was for hard work of manual operations after scram or LOCA accidents. An integrating system with the normal and the abnormal automatic systems are being developed for interacting smoothly both systems. (author)

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

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.

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

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...

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

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."

Reactor control system. PWR

International Nuclear Information System (INIS)

2009-01-01

At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

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

Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

Energy Technology Data Exchange (ETDEWEB)

Wang, C.-D. [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China); Nuclear Engineering Division, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)], E-mail: jdwang@iner.gov.tw; Lin Chaung [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China)

2009-08-15

An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.

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)

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

Computer-aided testing and operational aids for PARR-1 nuclear reactor

International Nuclear Information System (INIS)

Ansari, S.A.

1990-01-01

The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

Measurements of kinetic parameters by noise techniques on the MINERVE reactor

International Nuclear Information System (INIS)

Carre, J.C.; Da Costa Oliveira, J.

1975-01-01

Noise measurements were determined on ERMINE a fast thermal coupled reactor built in MINERVE. A reactor without feedback, and a reactor with an automatic control rod were both considered. The first case concerned the measurements of auto and cross power spectral density obtained with one or two neutron detectors, and the determination of: neutron lifetime; efficiency for one ion chamber; power level of the reactor; maximal speed and acceleration of the control rod for the design of an automatic reactor control actuator. The second case was concerned with measurements of the auto power spectral density in reactivity for the control rod, and the estimation of: the transfer function of the automatic pilot; the neutron lifetime; and the standard error affecting the results obtained by the oscillation method. The results proved that the pile noise theory with a point kinetic model is sufficient for application on zero power reactors. (U.K.)

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.

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

Feedback control of primary pump using midplane temperature of lower density lock for a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, Kanji; Haga, Katsuhiro; Tamaki, Masayoshi

1993-01-01

A new automatic pump speed control system, using a measurement of the temperature distribution in the lower density lock, is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the lower density lock in order to prevent the poison water from penetrating into the core during normal operation. In a startup test, the effectiveness of this control system to bring the system quickly to the stable state from a very small initial temperature difference between top and bottom of the lower density lock has been confirmed. The effectiveness of the primary pump trip at the limit speed in the control system to shutdown the core power safely in an accident such as a loss-of-feedwater accident with and without the primary loop isolation has also been proved

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.

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)

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

Digital implementation of AMSACs at Harris and Robinson plants

International Nuclear Information System (INIS)

Burjorjee, D.; Stepps, D.

1988-01-01

The Code of Federal Regulations was altered in July 1984 to include a section on Requirements for Reduction of Risk from Anticipated Transients Without Scram Events for Light Water Cooled Nuclear Power Plants. For pressurized water reactors the code required equipment diverse from the reactor trip system to automatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine trip under conditions indicative of an anticipated transient without scram (ATWS). The equipment in question is called ATWS mitigation system actuation circuitry (AMSAC). The AMSACs for Carolina Power and Light Company's Shearon Harris and Robinson power plants have been designed and built by Atomic Energy of Canada Limited (AECL) from commercially available components to meet stringent reliability requirements and minimize operational burdens

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

Reactor wall in thermonuclear device

International Nuclear Information System (INIS)

Shibui, Masanao.

1988-01-01

Purpose: To always monitor the life of armours in reactor walls and automatically shutdown the reactor if it should be operated in excess of the limit of use. Constitution: Monitoring material of lower melting point than armours (for example beryllium pellets) as one of the reactor wall constituents of a thermonuclear device are embedded in a region leaving the thickness corresponding to the allowable abrasion of the armour. In this structure, if the armours are abrased due to particle loads of a plasma and the abrasion exceeds a predetermined allowable level, the monitoring material is exposed to the plasma and melted and evaporated. Since this can be detected by impurity monitors disposed in the reactor, it is possible to recognize the limit for the working life of the armours. If the thermonuclear reactor should be operated accidentally exceeding the life of the armours, since a great amount of the monitoring materials have been evaporated, they flow into the plasma to increase the plasma radiation loss thereby automatically eliminate the plasma. (K.M.)

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.

Development of an automatic prompt gamma-ray activation analysis system

International Nuclear Information System (INIS)

Osawa, Takahito

2013-01-01

An automatic prompt gamma-ray activation analysis system was developed and installed at the Japan Research Reactor No. 3 Modified (JRR-3M). The main control software, referred to as AutoPGA, was developed using LabVIEW 2011 and the hand-made program can control all functions of the analytical system. The core of the new system is an automatic sample exchanger and measurement system with several additional automatic control functions integrated into the system. Up to fourteen samples can be automatically measured by the system. (author)

Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor; Diseno y construccion de un sistema electronico automatico de medicion y graficado del flujo neutronico para el reactor subcritico

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1997-07-01

The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

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

Automatic scheduling of maintenance work in nuclear power plants

International Nuclear Information System (INIS)

Kasahara, T.; Nishizawa, Y.; Kato, K.; Kiguchi, T.

1987-01-01

An automatic scheduling method for maintenance work in nuclear power plants has been developed using an AI technique. The purpose of this method is to help plant operators by adjusting the time schedule of various kinds of maintenance work so that incorrect ordering or timing of plant manipulations does not cause undersirable results, such as a plant trip. The functions of the method were tested by off-line simulations. The results show that the method can produce a satisfactory schedule of plant component manipulations without interference between the tasks and plant conditions

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

The FieldTrip-SimBio pipeline for EEG forward solutions.

Science.gov (United States)

Vorwerk, Johannes; Oostenveld, Robert; Piastra, Maria Carla; Magyari, Lilla; Wolters, Carsten H

2018-03-27

Accurately solving the electroencephalography (EEG) forward problem is crucial for precise EEG source analysis. Previous studies have shown that the use of multicompartment head models in combination with the finite element method (FEM) can yield high accuracies both numerically and with regard to the geometrical approximation of the human head. However, the workload for the generation of multicompartment head models has often been too high and the use of publicly available FEM implementations too complicated for a wider application of FEM in research studies. In this paper, we present a MATLAB-based pipeline that aims to resolve this lack of easy-to-use integrated software solutions. The presented pipeline allows for the easy application of five-compartment head models with the FEM within the FieldTrip toolbox for EEG source analysis. The FEM from the SimBio toolbox, more specifically the St. Venant approach, was integrated into the FieldTrip toolbox. We give a short sketch of the implementation and its application, and we perform a source localization of somatosensory evoked potentials (SEPs) using this pipeline. We then evaluate the accuracy that can be achieved using the automatically generated five-compartment hexahedral head model [skin, skull, cerebrospinal fluid (CSF), gray matter, white matter] in comparison to a highly accurate tetrahedral head model that was generated on the basis of a semiautomatic segmentation with very careful and time-consuming manual corrections. The source analysis of the SEP data correctly localizes the P20 component and achieves a high goodness of fit. The subsequent comparison to the highly detailed tetrahedral head model shows that the automatically generated five-compartment head model performs about as well as a highly detailed four-compartment head model (skin, skull, CSF, brain). This is a significant improvement in comparison to a three-compartment head model, which is frequently used in praxis, since the importance of

Release of fission products in transients

International Nuclear Information System (INIS)

Christensen, H.; Lundqwist, R.

1979-07-01

A station for automatic sampling of coolant has been put in operation at the Oskarshamn-1 reactor. The release of 131 J and other fission products in spikes in connection with reactor trips and scheduled shutdowns has been measured. A model developed at General Electric has been used to predict the spike release in Oskarshamn-1 and the predicted values have been compared with experimental values. Literature data of iodine spikes in BWR and PWR have been reviewed. (author)

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

Fuzzy power control algorithm for a pressurized water reactor

International Nuclear Information System (INIS)

Hah, Y.J.; Lee, B.W.

1994-01-01

A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rods with a conventional proportional-integral-differential controller because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability needed for load-following operations including frequency control. In an attempt to achieve automatic power shape control without any design modifications to the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multiple-input multiple-output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of PWRs during the load-following operations

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)

Automatic optimization of constants and special mathematic ensuring algorithms SKALA-micro system of RBMK-1000 reactor self-certification in operation

International Nuclear Information System (INIS)

Aleksandrov, S.I.; Dmitrenko, V.V.; Postnikov, V.V.; Sviridenkov, A.N.; Yurkin, G.V.; Yakunin, I.S.

2007-01-01

Paper dwells upon problems dealing with accuracy improvement of the energy release distribution and the safety margin of the RBMK-1000 operation. The accuracy is improved through the automatic optimization of some constants used in the SKALA-micro system special mathematic ensuring program and the regular self-validation of the algorithm to determine the energy release distribution calculation error. The validation based on the regular scanning of the reactor core by a calibrating detector and through the sequence disabling of the internal detectors is shown to give the close results [ru

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…

Vent control device for nuclear reactor container

International Nuclear Information System (INIS)

Kubota, Ryuji.

1989-01-01

The present invention concerns automatic prevention of abnormal over-pressure and hydrogen gas flashing in a BWR type reactor container. That is, (1) if the pressure in the container is abnormally increased, the gas in the pressure suppression chamber is released to reduce the pressure thereby preventing over-pressure damage to the container. (2) Then, if exhaust gases are burnt to cause flashing explosion danger for the gases in the reactor container, the gas release is interrupted. The foregoing two functioins are automatically conducted in this device. Specifically, when the pressure in the reactor container reaches a predetermined allowable limit, a remote control operation valve is opened by automatic control means to release the gas in the vessel. Since the gas flow rate at the start of the release exceeds flame propagation velocity, there is no worry for flashing explosion. Further, if the pipeway flow velocity near the atmospheric release is reduced to less than the flame propagation velocity of the hydrogen gas, the opened valve is automatically closed. Accordingly, propagation of hydrogen gas flame into the container thus causing explosion can surely be prevented. (K.M.)

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.

Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

Energy Technology Data Exchange (ETDEWEB)

Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2012-10-15

The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

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)

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.

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.

Neutron flux monitoring with pre-startup channels in FBTR [Paper No.:E6

International Nuclear Information System (INIS)

Nagaraj, C.P.; Ramakrishnan, R.; Subha Rao, R.; Pillai, C.P.; Muralikrishan, G.; Raghavan, K.

1993-01-01

The pre-start up instrumentation system using boron lined proportional counters has been well designed and the counters have been rigorously tested. This system in conjunction with the regular start-up channels (with more than 3 decades overlap) covers the start up measurement range adequately and provides necessary indications, interlocks and trips on low count rate, high count rate and doubling time to ensure safety. The pre-startup channels have been operational. The reactor trip on LOG NO due to count rate on start-up channels less than 3 cps is automatically inhibited with the higher count rate obtained on the pre-startup channels. With this and log CRM readings from PSU well on scale has facilitated smooth and safe reactor start ups even with low shutdown count rates. (author). 5 figs

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.

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.

Reactor power control systems in nuclear power plants

International Nuclear Information System (INIS)

Nakajima, Kazuo.

1980-01-01

Purpose: To enable power control by automatic control rod operation based on the calculated amounts of operation for the control rods determined depending on a power set value from reactor operators or on power variation amounts from other devices. Constitution: When an operator designates an automatic selection by way of a control rod operation panel, automatic signals are applied to a manual-automatic switching circuit and the mode judging circuit of a rod pattern control device. Then, mode signals such as for single operation, load setting, load following and the like produced by the operator are judged in a circuit, wherein a control rod pattern operation circuit calculates the designation for the control rods and the operation amounts for the control rods depending on the designated modes and automatic control is conducted for the control rods by a rod position control circuit, a rod drive control device and the like connected at a rod position monitor device. The reactor power is thus controlled automatically to reduce the operator's labours. The automatic power control can also be conducted in the same manner by the amount of power variations applied to the device from the external device. (Yoshino, Y.)

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

The design of control algorithm for automatic start-up model of HWRR

International Nuclear Information System (INIS)

Guo Wenqi

1990-01-01

The design of control algorithm for automatic start-up model of HWRR (Heavy Water Research Reactor), the calculation of μ value and the application of digital compensator are described. Finally The flow diagram of the automatic start-up and digital compensator program for HWRR are given

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)

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.

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.

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)

Automatization of the radiation control measurements

International Nuclear Information System (INIS)

Seki, Akio; Ogata, Harumi; Horikoshi, Yoshinori; Shirai, Kenji

1988-01-01

Plutonium Fuel Production Facility (PFPF) was constructed to fabricate the MOX fuels for 'MONJU' and 'JOYO' reactors and to develop the practical fuel fabricating technology. For the fuel fabrication process in this facility, centralized controlling system is being adopted for the mass production of the fuel and reduction of the radiation exposure dose. Also, the radiation control systems are suitable for the large-scale facility and the automatic-remote process of the fuel fabrication. One of the typical radiation control systems is the self moving survey system which has been developed by PNC and adopted for the automatic routine monitoring. (author)

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.

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.

Power conditioning system for a nuclear reactor

International Nuclear Information System (INIS)

Higashigawa, Yuichi; Joge, Toshio.

1981-01-01

Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

Parallelization and automatic data distribution for nuclear reactor simulations

Energy Technology Data Exchange (ETDEWEB)

Liebrock, L.M. [Liebrock-Hicks Research, Calumet, MI (United States)

1997-07-01

Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.

Parallelization and automatic data distribution for nuclear reactor simulations

International Nuclear Information System (INIS)

Liebrock, L.M.

1997-01-01

Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed

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.

Automatic radiometric analyzer for nuclides in nuclear reactor water

International Nuclear Information System (INIS)

Kitamura, Masao; Tokoi, Hiromi; Kitaguchi, Hiroshi; Ozawa, Yoshihiro; Urata, Megumu.

1981-01-01

Purpose: To shorten the processing time and improve the accuracy for processing water sampled from reactor coolants, as well as simplify the mechanism of the apparatus. Constitution: Reactor water sampled from reactor coolants, after filtered out with insoluble solids, is stored in an ion exchange container. Thereafter, the amount of ion exchanged water is regulated by the coarse measurement of radioactivity concentration by a monitor. Further, ion exchange resins are charged from a resin tank, agitated by gases and dispersed into sampled water. Then, all of the radioactive iodines contained in the sample are collected in the resins. The resins are recovered through evacuation into instrumenting vessels for measurement of radioactivity. Since ion exchange resins are dispersed in the sampled water in this system, the processing time can be shortened. (Ikeda, J.)

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)

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)

Development of new techniques and enhancement of automatic capability of neutron activation analysis at the Dalat Research Reactor

International Nuclear Information System (INIS)

Ho Manh Dung; Ho Van Doanh; Tran Quang Thien; Pham Ngoc Tuan; Pham Ngoc Son; Tran Quoc Duong; Nguyen Van Cuong; Nguyen Minh Tuan; Nguyen Giang; Nguyen Thi Sy

2017-01-01

The techniques of neutron activation analysis (NAA) including cyclic, epithermal and prompt-gamma (CNAA, ENAA and PGNAA, respectively) have been developed at the Dalat research reactor (DRR). In addition, the efforts has been spent to improve the automatic capability of irradiation, measurement and data processing of NAA. The renewal of necessary devices/tools for sample preparation have also been done. Eventually, the performance and the utility in terms of sensitivity, accuracy and stability of the analytical results generated by NAA at DRR have significantly been improved. The main results of the project are: 1) Upgrading of the fast irradiation system on Channel 13-2/TC to allow the cyclic irradiations; 2) Development of CNAA; 3) Development of ENAA; 4) Application of k0-method for PGNAA; 5) Investigation of the automatic sample changer (ASC2); 6) Upgrading of Ko-DALAT software for ENAA and modification of k0-IAEA software for CNAA and PGNAA; and 7) Optimization of irradiation and measurement facilities as well as sample preparation devices/tools. A set of procedures of relevant developed techniques in the project were established. The procedures have been evaluated by analysis of the reference materials for which they are meeting the requirements of multi-element analysis for the intended applications. (author)

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)

Plasma automatic control in magnetic traps

International Nuclear Information System (INIS)

Samojlenko, Yu.I.; Chuyanov, V.A.

1983-01-01

Principles of constructing the systems providing a plasma equilibrium and stability in thermonuctear devices are laid down. Operation of the servo system to maintain a plasma equilibrium is described using the tokamak plasma filament as an example. Operation of the system to suppress a flute instability is also described. This system measures electric disturbances on the plasma body surface and controls charge distribution on external electrodes. It is pointed out that systems of automatic control of plasma equilibrium and stability become an essential element of a future thermonuclear reactor and the system potentialities would much determine the reactor economic efficiency

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

Automatic radioxenon analyzer for CTBT monitoring

International Nuclear Information System (INIS)

Bowyer, T.W.; Abel, K.H.; Hensley, W.K.

1996-12-01

Over the past 3 years, with support from US DOE's NN-20 Comprehensive Test Ban Treaty (CTBT) R ampersand D program, PNNL has developed and demonstrated a fully automatic analyzer for collecting and measuring the four Xe radionuclides, 131m Xe(11.9 d), 133m Xe(2.19 d), 133 Xe (5.24 d), and 135 Xe(9.10 h), in the atmosphere. These radionuclides are important signatures in monitoring for compliance to a CTBT. Activity ratios permit discriminating radioxenon from nuclear detonation and that from nuclear reactor operations, nuclear fuel reprocessing, or medical isotope production and usage. In the analyzer, Xe is continuously and automatically separated from the atmosphere at flow rates of about 7 m 3 /h on sorption bed. Aliquots collected for 6-12 h are automatically analyzed by electron-photon coincidence spectrometry to produce sensitivities in the range of 20-100 μBq/m 3 of air, about 100-fold better than with reported laboratory-based procedures for short time collection intervals. Spectral data are automatically analyzed and the calculated radioxenon concentrations and raw gamma- ray spectra automatically transmitted to data centers

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.

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.

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)

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

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

Study of the self-regulating properties of a WWER reactor

International Nuclear Information System (INIS)

Filo, J.; Trnkusz, J.; Polak, V.

1979-01-01

The results of a self-regulation experiment carried out on the V-1 reactor in Czechoslovakia in the period of the start-up are presented. The kinetic state of the reactor was modified by varying the position of the automatic control rods, the power of the turbogenerators and by switching off a main pump in the primary circuit, on power levels of 35, 55, 75 and 90%. The important thermal parameters of the reactor, the electric power of the turbogenerators, the neutron flux, the position of the automatic control rod group and the concentration of the boric acid in the coolant have been measured. (R.J.)

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.

NEPTUNE: a modular system for light-water reactor calculation

International Nuclear Information System (INIS)

Bouchard, J.; Kanevoky, A.; Reuss, P.

1975-01-01

A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

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.

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...

Completely automated nuclear reactors for long-term operation

International Nuclear Information System (INIS)

Teller, E.; Ishikawa, M.; Wood, L.

1996-01-01

The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant's operational and post-operational life. These reactors are proposed to be situated in suitable environments at ∼100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant's heat engine and electrical generator subsystems are located above-ground

Developing remote techniques for liquid metal reactors

International Nuclear Information System (INIS)

Fenemore, Peter

1987-01-01

Three devices have been designed in Britain to meet the need for special remote equipment and techniques required to inspect the reactor vessel and internals of liquid metal reactors. The ''Links Manipulator Under-Sodium Viewing System'' - a device to be used for the surveillance of reactor internals, which are submerged in sodium. An ''Automatic Guided Vehicle'' - a free roving vehicle to be used to survey the externals of the reactor vessel. The ''Snake Manipulator'' - an articulated arm used to gain access to restricted areas. (author)

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.

Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

International Nuclear Information System (INIS)

Hanson, G.H.

1978-01-01

The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. A policy involves disabling the automatic-actuation of the LOFT ECCS, but still retaining the manual actuation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 33 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. For the operating power of the L2-2 Experiment, the required decay-periods (with operating periods of 40 and 2000 hours) are about 21 and 389 hours, respectively. With operating periods of 40 and 2000 hours at Core-I full power, the required decay-periods are about 42 and 973 hours, respectively. After these decay periods the automatic actuation of the LOFT ECCS can be disabled assuming a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. The automatic and manual lineup of the ECCS may be waived if decay power is less than 11 kW

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...

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.

Old and new ways in reactor technology. Reactor concepts and reactor safety

Energy Technology Data Exchange (ETDEWEB)

Schulten, R

1989-01-01

Compared to developments of other technical-scale systems, the period between the recognition of the underlying physics of nuclear fission and the development of a functioning nuclear reactor and its further development to the present level of maturity has been relatively short. The whole development is based on the chain reaction and is rendered safe by the possible auto-stabilization of this reaction. Consequently, the safety of nuclear reactors properly designed is based on automatic mechanisms, which prevent spreads of radioactivity even in major accidents. Controversial opinions about nuclear power uses are mostly based on wrong perceptions both of reactor safety and of radioactive waste, unless they are characterized by sheer ideology. The use of nuclear power worldwide has assumed an important, growing role in the combined uses of a variety energy sources in a surprisingly short period of time and will continue to make a safe, economic, and thus responsible contribution in the long run.

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)

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.

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.

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.

WRAP: a water reactor analysis package

International Nuclear Information System (INIS)

Anderson, M.M.

1977-06-01

The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

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...

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.

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...

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…

Computer program for automatic generation of BWR control rod patterns

International Nuclear Information System (INIS)

Taner, M.S.; Levine, S.H.; Hsia, M.Y.

1990-01-01

A computer program named OCTOPUS has been developed to automatically determine a control rod pattern that approximates some desired target power distribution as closely as possible without violating any thermal safety or reactor criticality constraints. The program OCTOPUS performs a semi-optimization task based on the method of approximation programming (MAP) to develop control rod patterns. The SIMULATE-E code is used to determine the nucleonic characteristics of the reactor core state

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)

Centralized digital computer control of a research nuclear reactor

International Nuclear Information System (INIS)

Crawford, K.C.

1987-01-01

A hardware and software design for the centralized control of a research nuclear reactor by a digital computer are presented, as well as an investigation of automatic-feedback control. Current reactor-control philosophies including redundancy, inherent safety in failure, and conservative-yet-operational scram initiation were used as the bases of the design. The control philosophies were applied to the power-monitoring system, the fuel-temperature monitoring system, the area-radiation monitoring system, and the overall system interaction. Unlike the single-function analog computers currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control-rod movements to conform with operator requests, automatically log the required physical parameters during reactor operation, perform the required system tests, and monitor facility safety and security. Reactor power control is based on signals received from ion chambers located near the reactor core. Absorber-rod movements are made to control the rate of power increase or decrease during power changes and to control the power level during steady-state operation. Additionally, the system incorporates a rudimentary level of artificial intelligence

Operating experience of the automatic technological control system at the Kolsk NPP

International Nuclear Information System (INIS)

Volkov, A.P.; Ignatenko, E.I.; Kolomtsev, Yu.V.; Mel'nikov, E.F.; Trofimov, B.A.

1981-01-01

Briefly reviewed is operating experience of the automatic control systems of the kolsk NPP (KNPP) power units, where measuring technique of the neutron flux ''Iney'', ARM-4 power regulator, automatic turbine start-up system ATS are used. The main shortcomings of the technological process automatic control system (ACS) and ways of their removal are considered. It is noted that the KNPP ACS performs only limited start-up functions of the basic equipment and reactor power control as well as partially protection functions at instant loading drops and switch-off of the main circulating pump [ru

Considerations concerning the reliability of reactor safety equipment

International Nuclear Information System (INIS)

Furet, J.; Guyot, Ch.

1967-01-01

A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [fr

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.

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)

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

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

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

Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

International Nuclear Information System (INIS)

Hanson, G.H.

1978-01-01

The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manual activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS

Level controlling system in BWR type reactors

International Nuclear Information System (INIS)

Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

1981-01-01

Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

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.

Development of a system for automatic control and performance evaluation of shutoff units in HANARO

International Nuclear Information System (INIS)

Jeong, Y. H.; Joe, Y. G.; Choi, Y. S.; Woo, J. S.

2003-01-01

The function of the shutoff units is to rapidly insert the shutoff rod into the reactor core for safe shutdown of reactor. This paper describes the development of a system for automatic control and performance evaluation of shutoff units. The system automatically drives the shutoff unit with a specified operation cycle and records the performance of the drive mechanism in graphs and data. Also, it records the operating parameters of the shutoff unit and test facility. The characteristic of the developed system was evaluated to compare with that being use in the HANARO reactor. The system will be used for the performance and endurance tests in the test facility. Hereafter, the system will efficiently be used for the normal operation and the periodical drop performance tests of shutoff units in HANARO

Automatic acoustic and vibration monitoring system for nuclear power plants

International Nuclear Information System (INIS)

Tothmatyas, Istvan; Illenyi, Andras; Kiss, Jozsef; Komaromi, Tibor; Nagy, Istvan; Olchvary, Geza

1990-01-01

A diagnostic system for nuclear power plant monitoring is described. Acoustic and vibration diagnostics can be applied to monitor various reactor components and auxiliary equipment including primary circuit machinery, leak detection, integrity of reactor vessel, loose parts monitoring. A noise diagnostic system has been developed for the Paks Nuclear Power Plant, to supervise the vibration state of primary circuit machinery. An automatic data acquisition and processing system is described for digitalizing and analysing diagnostic signals. (R.P.) 3 figs

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

Pressurised water reactor fuel management using PANTHER

International Nuclear Information System (INIS)

Parks, G.T.; Knight, M.P.

1996-01-01

This paper describes the integration of Nuclear Electric's reactor physics code PANTHER with an automatic optimisation procedure designed to search for optimal PWR reload cores and assesses its performance. (Author)

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

Method and apparatus for controlling the neutron flux in nuclear reactors

International Nuclear Information System (INIS)

Minnick, L.E.

1979-01-01

A control rod assembly in a nuclear reactor that automatically scrams the reactor when a loss of coolant flow occurs and that can also control the level of neutron flux in the reactor is described. The control rod assembly includes a separator plate having an orifice through which the reactor coolant flows and a sealing surface around the orifice. The control rod in the assembly has a complementary sealing surface. When the control rod and separator plate are brought into contact, the differential pressure across the separator plate caused by the flow of the primary coolant through the reactor core retains the two sealing surfaces together. If the flow of coolant stops or the differential pressure across the separator plate decreases for any reason, the control rod drops by gravity and the reactor is scrammed. The control rod is also automatically dropped as a result of the lateral vibration of an earthquake or by the downward motion of the rod drive shaft, either of which will open the sealing surfaces and reduce the sealing pressure

Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

International Nuclear Information System (INIS)

Boeck, H.; Weiss, H.; Hood, W.E.; Hyde, W.K.

1992-01-01

The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert is assigned the responsibility of reviewing the proposed installation and verifying all safety aspects. After a positive review, final assembly and checkout of the instrumentation system may commence. The instrumentation system consists of three basic modules: the control system console, the data acquisition console and the NH-1000 wide range channel. Digital communications greatly reduce interwiring requirements. Hardwired safety channels are independent of computer control, thus, the instrumentation system in no way relies on any computer intervention for safety function. In addition, both the CSC and DAC computers are continuously monitored for proper operation via watchdog circuits which are capable of shutting down the reactor in the event of computer malfunction. Safety channels include two interlocked NMP-1000 multi-range linear channels for steady state mode, an NPP-1000 linear safety channel for pulse mode and a set of three independent fuel temperature monitoring channels. The microprocessor controlled wide range NM- 1000 digital neutron monitor (fission chamber based) functions as a startup/operational channel, and provides all power level related Interlocks. The Atominstitut TRIGA reactor is configured for four modes of operation: manual mode, automatic mode (servo control), pulsing mode and square wave mode. Control of the standard control rods is via stepping motor control rod drives, which offers the operator the choice of which control rods are operated by the servo system in automatic and square wave model. (author)

Heating control system for nuclear reactor

International Nuclear Information System (INIS)

Shinohara, Kaoru.

1981-01-01

Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

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

Digital computer control of a research nuclear reactor

International Nuclear Information System (INIS)

Crawford, Kevan

1986-01-01

Currently, the use of digital computers in energy producing systems has been limited to data acquisition functions. These computers have greatly reduced human involvement in the moment to moment decision process and the crisis decision process, thereby improving the safety of the dynamic energy producing systems. However, in addition to data acquisition, control of energy producing systems also includes data comparison, decision making, and control actions. The majority of the later functions are accomplished through the use of analog computers in a distributed configuration. The lack of cooperation and hence, inefficiency in distributed control, and the extent of human interaction in critical phases of control have provided the incentive to improve the later three functions of energy systems control. Properly applied, centralized control by digital computers can increase efficiency by making the system react as a single unit and by implementing efficient power changes to match demand. Additionally, safety will be improved by further limiting human involvement to action only in the case of a failure of the centralized control system. This paper presents a hardware and software design for the centralized control of a research nuclear reactor by a digital computer. Current nuclear reactor control philosophies which include redundancy, inherent safety in failure, and conservative yet operational scram initiation were used as the bases of the design. The control philosophies were applied to the power monitoring system, the fuel temperature monitoring system, the area radiation monitoring system, and the overall system interaction. Unlike the single function analog computers that are currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control rod movements to conform with operator requests, automatically log the required physical parameters during reactor

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)

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...

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

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...

Computer simulation system of neural PID control on nuclear reactor

International Nuclear Information System (INIS)

Chen Yuzhong; Yang Kaijun; Shen Yongping

2001-01-01

Neural network proportional integral differential (PID) controller on nuclear reactor is designed, and the control process is simulated by computer. The simulation result show that neutral network PID controller can automatically adjust its parameter to ideal state, and good control result can be gotten in reactor control process

Application of pattern recognition technique on randon signals for automatic monitoring of dynamic systems with emphasis on nuclear reactors

International Nuclear Information System (INIS)

Nascimento, J.A. do.

1981-01-01

The time varying or noise component of dynamic system parameters contains information on the system state. Pattern recognition analysis of noise signals for such systems is a powerful technique for assessing 'system normality' or 'correct operation'. Data analysis with modern small computers enables the otherwise unmanageable volumes of data to be processed on line and the results presented in a meaningful form. These informations provide necessary data for maintaining the system at optimum operating conditions. An automatic pattern recognition program, PSDREC, developmed for the surveillance of nuclear reactor and rotating machinery is described, and the relevant theory is outlined. This program, which applies 8 statistical tests to calculated power spectral density (PSD) distributions, was earlier installed in a PDP-11/45 computer at IPEN. In this work it has been used to separately analyse recorded signals from three systems, namely an operational BWR power reactor (neutron signals), a water pump and a diesel motor (vibration signals). The latter two were, respectively, operated over a wide-range of flow and load conditions. The statistical tests were applied to frequency bands of (0,1-40) Hz, (0-1000) Hz and (0,20000) Hz. for the BWR, pump and diesel signal data, respectively. Operation and analysis conditions are given together with representative graphs of the analysed PSD distributions. Results of the tests - discussed in some detail - are considered to be satisfactory. (Author) [pt

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

Pneumatic transport systems for TRIGA reactors

International Nuclear Information System (INIS)

Bolton, John A.

1970-01-01

Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

Operational experience of the Marcoule reactors

International Nuclear Information System (INIS)

Conte, F.

1963-01-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

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

ATWS analyses. Analysis of anticipated transients without reactor scram in Combustion Engineering NSSS's

International Nuclear Information System (INIS)

1976-05-01

Results are presented of analyses of the transient thermal-hydraulic conditions and radiological release consequences which would occur in power plants which employ a Combustion Engineering Nuclear Steam Supply System during Anticipated Transients Without Scram due to a lack of insertion of the Control Element Assemblies upon signals for automatic or manual reactor shutdown. The transients analyzed include all events which meet the criterion to be considered as anticipated at least once in the plant lifetime with automatic reactor shutdown

Small reactor operating mode

International Nuclear Information System (INIS)

Snell, V.G.

1997-01-01

There is a potential need for small reactors in the future for applications such as district heating, electricity production at remote sites, and desalination. Nuclear power can provide these at low cost and with insignificant pollution. The economies required by the small scale application, and/or the remote location, require a review of the size and location of the operating staff. Current concepts range all the way from reactors which are fully automatic, and need no local attention for days or weeks, to those with reduced local staff. In general the less dependent a reactor is on local human intervention, the greater its dependence on intrinsic safety features such as passive decay heat removal, low-stored energy and limited reactivity speed and depth in the control systems. A case study of the design and licensing of the SLOWPOKE Energy System heating reactor is presented. (author)

High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

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...

Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor

International Nuclear Information System (INIS)

Polo-López, M.I.; Fernández-Ibáñez, P.; Ubomba-Jaswa, E.; Navntoft, C.; García-Fernández, I.; Dunlop, P.S.M.; Schmid, M.; Byrne, J.A.

2011-01-01

Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor

Energy Technology Data Exchange (ETDEWEB)

Polo-Lopez, M.I., E-mail: mpolo@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Fernandez-Ibanez, P., E-mail: pilar.fernandez@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Ubomba-Jaswa, E., E-mail: euniceubombajaswa@yahoo.com [Natural Resources and the Environment, CSIR, PO Box 395, Pretoria (South Africa); Navntoft, C., E-mail: christian.navntoft@solarmate.com.ar [Instituto de Investigacion e Ingenieria Ambiental, Universidad Nacional de San Martin (3iA-UNSAM), Peatonal Belgrano 3563, B1650ANQ San Martin (Argentina); Universidad Tecnologica Nacional - Facultad Regional Buenos Aires - Departamento de Ingenieria Civil - Laboratorio de Estudios sobre Energia Solar, (UTN-FRBA-LESES), Mozart 2300, (1407) Ciudad Autonoma de Buenos Aires, Republica Argentina (Argentina); Garcia-Fernandez, I., E-mail: irene.garcia@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Dunlop, P.S.M., E-mail: psm.dunlop@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Schmid, M. [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Byrne, J.A., E-mail: j.byrne@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); and others

2011-11-30

Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

Simulation of the turbine trip of Unit 1 of the Laguna Verde nuclear power plant using the code Simulate-3K

International Nuclear Information System (INIS)

Alegria A, A.; Filio L, C.; Ortiz V, J.

2017-09-01

In order to compare the results obtained from the model developed in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) with the code Simulate-3K (S3K) with respect to those reported by the process computer of the Central (SIIP), the simulation of the turbine trip transient was carried out, caused by the firing of the main generator, the low differential pressure of oil of its seals and the automatic Scram of Unit 1 of the Laguna Verde nuclear power plant, at 87% of power nominal during the operation cycle 16. Since the reactor was brought to a safe stop due to Scram, was enough to simulate 20 seconds to observe the maximum increase in pressure with S3K. In this work, the following parameters are shown and compared: the neutron flux, the thermal power, the pressure in the dome, the flow at the entrance to the core, the steam flow that leaves the vessel and the minimal critical power ratio (MCPR). The neutron flux of the average power range monitors of the nuclear power plant was compared with the S3K detectors model. Finally, the MCPR was calculated with a different correlation to that of the fuel supplier and its deviation from its safety limit was determined. In conclusion, the results obtained show the current state of the model for the simulation of reactivity transients and the opportunity areas to consolidate this tool in support of the process of licensing refueling in the CNSNS. (Author)

Automatic fuel lattice design in a boiling water reactor using a particle swarm optimization algorithm and local search

International Nuclear Information System (INIS)

Lin Chaung; Lin, Tung-Hsien

2012-01-01

Highlights: ► The automatic procedure was developed to design the radial enrichment and gadolinia (Gd) distribution of fuel lattice. ► The method is based on a particle swarm optimization algorithm and local search. ► The design goal were to achieve the minimum local peaking factor. ► The number of fuel pins with Gd and Gd concentration are fixed to reduce search complexity. ► In this study, three axial sections are design and lattice performance is calculated using CASMO-4. - Abstract: The axial section of fuel assembly in a boiling water reactor (BWR) consists of five or six different distributions; this requires a radial lattice design. In this study, an automatic procedure based on a particle swarm optimization (PSO) algorithm and local search was developed to design the radial enrichment and gadolinia (Gd) distribution of the fuel lattice. The design goals were to achieve the minimum local peaking factor (LPF), and to come as close as possible to the specified target average enrichment and target infinite multiplication factor (k ∞ ), in which the number of fuel pins with Gd and Gd concentration are fixed. In this study, three axial sections are designed, and lattice performance is calculated using CASMO-4. Finally, the neutron cross section library of the designed lattice is established by CMSLINK; the core status during depletion, such as thermal limits, cold shutdown margin and cycle length, are then calculated using SIMULATE-3 in order to confirm that the lattice design satisfies the design requirements.

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)

EMERIS: an advanced information system for a materials testing reactor

International Nuclear Information System (INIS)

Adorjan, F.; Buerger, L.; Lux, I.; Mesko, L.; Szabo, K.; Vegh, J.; Ivanov, V.V.; Mozhaev, A.A.; Yakovlev, V.V.

1990-06-01

The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

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.

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)

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

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

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

Automatic diagnosis of multiple alarms for reactor-control rooms

International Nuclear Information System (INIS)

Gimmy, K.L.; Nomm, E.

1981-01-01

A system has been developed at the Savannah River Plant to help reactor operators respond to multiple alarms in a developing incident situation. The need for such systems has become evident in recent years, particularly after the three Mile Island incident

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

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

Nuclear reactor container

International Nuclear Information System (INIS)

Ishiyama, Takenori.

1989-01-01

This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

Complete automation of nuclear reactors control

International Nuclear Information System (INIS)

Weill, J.

1955-01-01

The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P max and R max (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P max is reached. (M.P.)

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)

An integrative conceptual framework for analyzing customer satisfaction with shopping trip experiences in grocery retailing

DEFF Research Database (Denmark)

Esbjerg, Lars; Jensen, Birger Boutrup; Bech-Larsen, Tino

2012-01-01

Grocery retailers aim to satisfy customers, and because grocery shopping trips are frequently recurring, they must do socontinuously. Surprisingly, little research has addressed satisfaction with individual grocery shopping trips. This article therefore develops a conceptual framework for analyzing...... customer satisfaction with individual grocery shopping trip experiences within a overall ‘disconfirmation of expectations model’ of customer satisfaction. The contribution of the framework is twofold. First, by focusing on satisfaction with individual grocery shopping trips, previous research...... on satisfaction in the retailing literature. Second, the framework synthesizes and integrates multiple central concepts from different research streams into a common framework for analyzing shopping trip satisfaction. Propositions are derived regarding the relationships among the different concepts...