Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

A customized digital monitoring and display system for nonpower reactors

International Nuclear Information System (INIS)

Ficaro, E.P.; Wehe, D.K.

1989-01-01

A digital data acquisition system for monitoring plant variables has been designed and implemented at the University of Michigan's Ford Nuclear Reactor (FNR), a 2-MW open-pool research reactor. The digital data provided by this system will be useful for: improved operator training, real-time experimental calculations, noise analysis, closed-loop control, and expert system applications. This paper describes the analog-to-digital (A/D) transitions and the associated applications and benefits experienced

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

Upgrading the Siemens Argonaut Reactor Graz with a digital monitoring system

International Nuclear Information System (INIS)

Froehlich, O.; Ninaus, W.

1999-01-01

This paper presents a modern design of a reactor monitoring system (MS) which was developed for a research reactor. This MS is using digital concepts, and it is more flexible than an analog MS, it co-operates better with the user, and it is a very helpful tool for a training-reactor in an university environment. The heart of the system is a process computer, and it was possible to access all important signals and functions of the original nuclear instrumentation by additional hardware. The monitoring software was written in C for the platform '32Bit-DOS-protected-mode' and shows on several high-resolution screen pages all the collected signals and the working conditions of the reactor. Moreover, all signals which are recorded on the random access memory can be saved to the hard disk of the computer and may thereby be used offline as well.(author)

The radon monitoring system in Daya Bay Reactor Neutrino Experiment

International Nuclear Information System (INIS)

Chu, M.C.; Kwan, K.K.; Kwok, M.W.; Kwok, T.; Leung, J.K.C.; Leung, K.Y.; Lin, Y.C.; Luk, K.B.; Pun, C.S.J.

2016-01-01

We developed a highly sensitive, reliable and portable automatic system (H 3 ) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H 3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m 3 . This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

On-line system for monitoring of boiling in nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Tuerkcan, E.; Kozma, R.; Verhoef, J.P.; Nabeshima, K.

1996-01-01

An important goal of nuclear reactor instrumentation is the continuous monitoring of the state of the reactor and the detection of deviations from the normal behaviour at an early stage. Early detection of anomalies enables one to make the necessary steps in order to prevent further damage of nuclear fuel. In the present paper, an on-line core monitoring system is described by means of which boiling anomaly in nuclear reactor fuel assemblies can be detected. (author). 9 refs, 7 figs

Automation of the radiation protection monitoring system in the RP-10 reactor

International Nuclear Information System (INIS)

Anaya G, Olgger; Castillo Y, Walter; Ovalle S, Edgar

2002-01-01

During the reactor operation, it is necessary to carry out the radiological control in the different places of the reactor, in periodic form and to take a registration of these values. For it the radioprotection official, makes every certain periods, settled down in the procedures, to verify and to carry out the registration of those values in manual form of each one of the radiation monitors. For this reason it was carried out the design and implementation of an automatic monitoring system of radioprotection in the reactor. In the development it has been considered the installation of a acquisition data system for 27 radiation gamma monitors of the type Geiger Mueller, installed inside the different places of the reactor and in the laboratories where they are manipulated radioactive material, using as hardware the FieldPoint for the possessing and digitalization of the signs which are correspondents using the communication protocol RS-232 to a PC in which has settled a program in graphic environment that has been developed using the tools of the programming software LabWindows/CVI. Then, these same signs are sent 'on line' to another PC that is in the Emergency Center of Coordination to 500 m of the reactor, by means of a system of radiofrequency communication. (author)

Monitoring nuclear reactor systems using neural networks and fuzzy logic

International Nuclear Information System (INIS)

Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E.; Mullens, J.A.

1991-01-01

A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN's provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such ''virtual measurements'' the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up or performance can be determined. In the methodology presented the output of a virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control valve of an experimental reactor using data obtained during a start-up. The enhanced noise tolerance of the methodology is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems. 8 refs., 11 figs., 1 tab

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

Development of Operational Safety Monitoring System and Emergency Preparedness Advisory System for CANDU Reactors (I)

International Nuclear Information System (INIS)

Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho; Son, Han Seong; Song, Deok Yong

2007-01-01

As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code

A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings.

Science.gov (United States)

Zhou, Jianguo; Xu, Yaming; Zhang, Tao

2016-06-14

Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes.

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)

Monitoring nuclear reactor systems using neural networks and fuzzy logic

International Nuclear Information System (INIS)

Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E.; Mullens, J.A.

1992-01-01

A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN's provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such virtual measurements the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up-or performance can be determined. In the methodology presented the output of virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems

Multiprocessor based data acquisition system for radiation monitoring in nuclear reactors

International Nuclear Information System (INIS)

Pansare, M.G.; Narsaiah, A.; Anantha Krishnan, T.S.

1989-01-01

Expensive minicomputers are required for building powerful Data Acquisition Systems (DAS) capable of scanning and processing large number of signals in a real-time environment. However by using the inexpensive microprocessors in multiprocessor configuration it is possible to build DASs that are as powerful as minicomputer based systems at much lesser cost. This paper describes such a multiprocessor based DAS designed for acquiring data from various radiation monitoring instruments of a nuclear reactor. The system is built by using MULTIBUS standard boards based on intel 8086, 16 bit microprocessor, with local and shared memory. The system monitors upto 128 analog input channels, 64 digital input channels and actuates upto 128 digital output contacts. The system continuously checks for the alarm condition of the input channels and displays the alarm status on an ALARM CRT. Facility has been provided for the transfer of data to a central computer. At any instant of time, the information regarding different channels being monitored is available from the local console as well as through five remote terminals located at various places in the reactor building. (author)

Post-accident monitoring systems in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

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

2018-01-01

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

Online monitoring and diagnostic system on RA-6 nuclear reactor

International Nuclear Information System (INIS)

Garcia Peyrano, O. A.; Marticorena, M.; Koch, R. G.; Martinez, J. S; Berruti, G. E.; Nunez, W. M.; Gonzales, L. A.; Tarquini, L. D.; Sotelo, J. P

2009-01-01

This paper presents the Online Automatic Monitoring and Diagnostic System for mechanical components, installed on RA-6 Nuclear Reactor (San Carlos de Bariloche, Argentina). This system has been designed, installed and set-up by the Vibrations and Mechatronics Laboratory (Centro Atomico Bariloche, Comision Nacional de Energia Atomica) and Sitrack.com Argentina SA. This system provides an online mechanical diagnostic of the main reactor components, allowing incipient failures to be early detected and identified, avoiding unscheduled shut-downs and reducing maintenance times. The diagnostic is accomplished by an online analysis of the vibratory signature of the mechanical components, obtained by vibrations sensors on the main pump and the decay tank. The mechanical diagnostic and the main operational parameters are displayed on the reactor control room and published on the internet. [es

Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

OpenAIRE

Verma, Vasudha

2017-01-01

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

Loose parts monitoring in light water reactor cooling systems

International Nuclear Information System (INIS)

Santos, A.; Alma, B.J.

1982-01-01

The work related to loose monitoring system for light water reactor, developed at GRS - Munique, are described. The basic problems due to the exact localization and detection of the loose part as well the research activities and development necessary aiming to obtain the best techniques in this field. (E.G.) [pt

Advanced core monitoring technology for WWER reactors

International Nuclear Information System (INIS)

Nguyen, T.Q.; Casadei, A.L.; Doshi, P.K.

1993-01-01

The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

International Nuclear Information System (INIS)

Stanc, S.; Repa, M.

2001-01-01

Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

Monitoring circuit for reactor safety systems

Science.gov (United States)

Keefe, Donald J.

1976-01-01

The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

Monitoring circuit for reactor safety systems

International Nuclear Information System (INIS)

Keefe, D.J.

1976-01-01

The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned. 3 claims, 2 figures

Power distribution monitoring system in the boiling water cooled reactor core

International Nuclear Information System (INIS)

Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

1987-01-01

Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

On-line fatigue monitoring system for reactor pressure vessel

International Nuclear Information System (INIS)

Tokunaga, K.; Sakai, A.; Aoki, T.; Ranganath, S.; Stevens, G.L.

1994-01-01

A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to an operating boiling water reactor (BWR), Tsuruga Unit-1, is described. The system uses the influence function approach and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computed fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification No.501. Fatigue usage results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension. (author)

Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

International Nuclear Information System (INIS)

Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

2010-01-01

Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

2009-01-01

The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

A personal computer based console monitor for a TRIGA reactor

International Nuclear Information System (INIS)

Rieke, Phillip E.; Hood, William E.; Razvi, Junaid

1990-01-01

Numerous improvements have been made to the Mark F facility to provide a minimum reactor down time, giving a high reactor availability. A program was undertaken to enhance the monitoring capabilities of the instrumentation and control system on this reactor. To that end, a personal computer based console monitoring system has been developed, installed in the control room and is operational to provide real-time monitoring and display of a variety of reactor operating parameters. This system is based on commercially available hardware and an applications software package developed internally at the GA facility. It has (a) assisted the operator in controlling reactor parameters to maintain the high degree of power stability required during extended runs with thermionic devices in-core, and (b) provided data trending and archiving capabilities on all monitored channels to allow a post-mortem analysis to be performed on any of the monitored parameters

Radioactivity Monitoring System for TRIGA 2000 Reactor Water Tank with On-Line Gamma Spectrometer

International Nuclear Information System (INIS)

Prasetyo Basuki; Sudjatmi KA

2009-01-01

One of the requirements in radiological safety in the operating condition of research reactor are the absence of radionuclide from fission product released to reactor cooling water and environment. Early detection of fission product that released from fuel element can be done by monitoring radioactivity level on primary cooling water.Reactor cooling water can be used as an important indicator in detecting radioactivity level of material fission product, when the leakage occurs. Therefore, it needs to build a monitoring system for measuring radioactivity level of cooling water directly and simple. The idea of this system is counting radioactivity water flow from reactor tank to the marinelli cube that attached to the HPGe detector on gamma spectrometer. Cooling water from tank aimed on plastic pipe to the marinelli cube. Water flows in gravitational driven to the marinelli cube, with volume flow rate 5.1 liters/minute in the inlet and 2.2 liters/minute in output. (author)

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)

The development of ex-core neutron flux monitoring system for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

The development of ex-core neutron flux monitoring system for integral reactor

International Nuclear Information System (INIS)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S.

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying the ex

Progress for on-line acoustic emission monitoring of cracks in reactor systems

International Nuclear Information System (INIS)

Hutton, P.H.; Friesel, M.A.; Kurtz, R.J.

1985-10-01

This paper reviews FY1985 accomplishments and FY1986 plans for the NRC sponsored research program concerned with ''Acoustic Emission/Flaw Relationships for Inservice Monitoring of Nuclear Reactor Pressure Boundaries''. The objective of the acoustic emission (AE) monitoring program is to develop and validate the use of AE methods for continuous surveillance of reactor pressure boundaries to detect flaw growth. Topics discussed include testing AE monitoring on reactors, refinement of an AE signal identification relationship, study of slow crack growth rate effects on AE generation, and activity to produce an ASTM standard for AE monitoring and to gain ASME code acceptance of AE monitoring

A microprocessor based monitoring system for a small nuclear reactor facility

International Nuclear Information System (INIS)

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

1980-01-01

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

Subcriticality monitoring method for reactor

International Nuclear Information System (INIS)

Ueda, Makoto.

1991-01-01

The present invention accurately monitors the reactor subcriticality and ensures the critical safety, irrespective of the presence or absence of artificial neutron sources. That is, when the subcriticality is monitored upon reactivity changing operation which causes reactivity change to the reactor during shutdown, neutron monitors are disposed at a plurality of monitoring positions. Then, neutron counting ratio before and after conducting the reactivity changing operation is determined. The subcriticality of the reactor is monitored by the ratio and the state of scattering of the ratio of neutron counting rate between each of the neutron monitors. With such procedures, signals of the neutron monitors are used, the characteristic that the change of the signals depend on the change of the neutron multiplication of the reactor core can be utilized whether artificial neutron sources (external neutron sources) are disposed or not. Accordingly, the subcriticality can be monitored more reliably. (I.S.)

Computer monitoring of the RB reactor operation

International Nuclear Information System (INIS)

Milovanovic, S.; Pesic, M.; Milovanovic, T.

1998-01-01

Personal computer based acquisition system designed for monitoring of operation of the RB experimental reactor in the Institute of Nuclear Sciences 'Vinca' (former 'Boris Kidric') and experiences acquired during its use are shown in this paper. The monitoring covers generally all nuclear aspects of the reactor operation (start-up, nominal power operation, power changing, shut down and maintenance), but the emphasis is put on: real time (especially fast changing) reactivity measurement; supervising time dependence of the safety rods positions during shut down, and detection of position inaccuracy or failure operation of safety/control rods during the reactor operation or maintenance. (author)

Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

1999-05-15

The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

Monitoring for shaft cracks on reactor recirculation pumps

International Nuclear Information System (INIS)

Kowal, M.G.; O'Brien, J.T. Jr.

1989-01-01

The article discusses the vibration characteristics associated with a boiling water reactor (BWR) recirculation pump. It also describes the application of diagnostic techniques and shaft crack theory to an on-line diagnostic monitoring system for reactor recirculation pumps employed at Philadelphia Electric Company's Peach Bottom Atomic Power Station. Specific emphasis is placed on the unique monitoring techniques associated with these variable speed vertical pumps

The system for diagnostics and monitoring of the IBR-2 reactor state. Data acquisition, accumulation and storage of information

International Nuclear Information System (INIS)

Ermilov, V.G.; Ivanov, V.V.; Korolev, V.S.; Pepelyshev, Yu.N.; Semashko, S.V.; Tulaev, A.B.

2000-01-01

The architectural decisions for a developed distributed system of the IBR-2 pulsed reactor conditions monitoring are described. The system is intended for measurement of the basic reactor parameters, acquisition, storage and processing of information, the current reactor state monitoring, analysis of reactor parameters for a long time operation period both in on-line, and in off-line modes. The system is constructed in the architecture client-server using DBMS MS SQL Server 7.0 The basic hardware components of the system are measuring workstations and devices, processing and user workstations and the central server. The software of the system consists of the measuring programs, data flows dispatching services, client applications for data processing and visualization, and means for preparing data for subsequent presentation in WWW. The basic results of the first system operation phase and prospect of its development are discussed. (author)

Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Malathi, N.; Sahoo, P., E-mail: sahoop@igcar.gov.in; Ananthanarayanan, R.; Murali, N. [Real Time Systems Division, Electronics, Instrumentation and Radiological Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2015-02-15

An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

Computer based core monitoring system for an operating CANDU reactor

International Nuclear Information System (INIS)

Yoon, Moon Young; Kwon, O Hwan; Kim, Kyung Hwa; Yeom, Choong Sub

2004-01-01

The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same(0.008%), which showed that the CCMS could monitor core behaviors well

A new reactor core monitoring system. First experience gained at the Dukovany NPP

International Nuclear Information System (INIS)

Pecka, M.; Svarny, J.; Kment, J.

2001-01-01

The article deals with methods of interpretation of in-core measurements that are based on the determination of the three-dimensional (3D) power distribution within the reactor core, discusses on-line mode calculations, and describes the results obtained during the trial operation of the new SCORPIO-VVER reactor core monitoring system. The principles of the method of determination of the fuel assembly subchannel parameters are outlined. Alternative methods of self-powered detector signal conversion to local power are given, and some results of their testing are presented. Emphasis is put on self-powered detectors supplied by the US firm IST, which were first deployed at the Dukovany NPP in 1998. The predictive function of the SCORPIO-VVER system, whose implementation was inspired by favourable experience gained on some PWR reactors (such as the products of the Halden reactor project at Ringhals and Sizewell B) were adapted to the specific needs of WWER-440 reactors. The main results of validation of the functions are described and presented in detail. (author)

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

Remote Reactor Monitoring

Energy Technology Data Exchange (ETDEWEB)

Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-21

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

Development of a 10-decade single-mode reactor flux monitoring system

International Nuclear Information System (INIS)

Valentine, K.H.; Shepard, R.L.; Falter, K.G.; Reese, W.B.

1988-01-01

Conventional wide-range neutron channels employ three optional modes to monitor the required flux range from source levels to full power (typically 10 or more decades). Difficult calibrations are necessary to provide a continuous output signal when such a system switches from counting mode in the source range to mean-square voltage mode in the midrange to dc current mode in the power range. In an ORNL proof-of-principle test, a method of extended range counting was implemented with a fission counter and conventional wide-band pulse processing electronics to provide a single-mode, monotonically increasing signal that spanned /approximately 10/ decades of neutron flux. Ongoing work includes design, fabrication, and testing of a comlpete neutron flux monitoring system suitable for advanced liquid metal reactor designs. 6 refs., 4 figs

Development of Vibration Diagnostic System in Research Reactors

International Nuclear Information System (INIS)

EL-Kafas, A. A.

1999-01-01

Early failure detection and diagnosis system are an important group with increasing interest with the operating support system. Already existing system to monitor integrity of primary system components are vibration and acoustic monitoring system (2,3). The development of vibration diagnostic system for MARIA reactor (30 MW)-the second research reactor in Poland -was made. The new system is applied for the Egypt research reactor (ETRR-1). This paper represents the result obtained during the operation of this activity that carried out at MARIA and ETRR-1 reactors

Experiment monitoring system of a new electromagnet drive for nuclear reactor control rod

International Nuclear Information System (INIS)

Zhang Jige; Wang Xiaoguang; Wu Yuanqiang; Zhang Zhengming

2003-01-01

In order to deal with some unsolved problems in the engineering prototype design of a new electromagnet drive device for nuclear reactor control rod, the property experiment in view of principle prototype is carried out. Actual displacement of nuclear reactor control rod is measured by means of raster ruler and the test data is obtained by means of computer. The computer communicates with PLC using RS232 serial port. The experimental results show that the monitoring system have the properties of high reliability and high precision, and ensures the experiment to accomplish successfully

Savannah River Plant remote environmental monitoring system

International Nuclear Information System (INIS)

Schubert, J.F.

1987-01-01

The SRP remote environmental monitoring system consists of separations facilities stack monitors, production reactor stack monitors, twelve site perimeter monitors, river and stream monitors, a geostationary operational environmental satellite (GOES) data link, reactor cooling lake thermal monitors, meteorological tower system, Weather Information and Display (WIND) system computer, and the VANTAGE data base management system. The remote environmental monitoring system when fully implemented will provide automatic monitoring of key stack releases and automatic inclusion of these source terms in the emergency response codes

Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

International Nuclear Information System (INIS)

Ricardo, Carla Pereira; Mesquita, Amir Zacarias

2007-01-01

During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

Monitoring and control of the Rossendorf research reactor using a microcomputerized automation system

International Nuclear Information System (INIS)

Ba weg, F.; Enkelmann, W.; Klebau, J.

1982-01-01

A decentral hierarchic information system (HIS) is presented, which has been developed for monitoring and control of the Rossendorf Research Reactor RFR, but which may also be considered the prototype of a digital automation system (AS) to be used in power stations. The functions integrated in the HIS are as follows: process monitoring, process control, and use of a specialized industrial robot for control of charging and discharging of the materials to be irradiated. The AS is realized on the basis of the process computer system PRA 30 (A 6492) developed in the GDR and including a computer K 1630 and the intelligent process terminals ursadat 5000 connected by a fast serial interface (IFLS). (author)

Study on Turn-to-turn Short Circuit On一line Monitoring System for Dry一type Ai r一core Reactor

Directory of Open Access Journals (Sweden)

GAO Zi-wei

2017-04-01

Full Text Available The change of current value caused by turn-to-turn short circuit of dry-type air-core reactor is so little that failure detection is difficult to be carried out. In order to solve this problem，a new on-line monitoring system based on impedance variation of turn-to-turn short circuit is proposed. The numerical method is applied to analyze the variation of equivalent resistance and equivalent reactance when dry-type air-core reactor winding short circuit happens in different places，and the monitoring method based on harmonic analysis method and quasi- synchronization sampling method is analyzed by theory. The hardware system，which takes single-chip microcomputer as the core of data processing and logic control，completes data acquisition of voltage signal and current signal of the reactor. In the respect of software design，the impedance variation will be uploaded to the PC after it has been calculated by using the above monitoring method，and then monitoring of turn-to-turn short circuit fault will be realized. Finally，the design of on-line monitoring system is studied by testing. The research result shows that，the equivalent resistance increases and the equivalent reactance decreases when turn-to-turn short circuit occurs，and the variation of equivalent resistance is more obvious than equivalent reactance. The experiment results prove that this monitoring method is true and the on-line monitoring system is feasible.

Acoustic Emission for on-line reactor pressure boundary monitoring

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

1985-01-01

The program objective is to develop AE for continuous surveillance to assess flaw growth in reactor pressure boundaries. Technology in the laboratory is being evaluated on structures. Results have demonstrated basic feasibility of the program objective. AE monitoring a long term fatigue test of a pressure vessel demonstrated an instrument system, and the ability to detect unexpected as well as well as known fatigue cracks. Monitoring a nuclear reactor system shows that the coolant flow noise problem is manageable and AE can be detected under simulated operating conditions

On-line Monitoring of Instrumentation in Research Reactors

International Nuclear Information System (INIS)

2017-12-01

This publication is the result of a benchmarking effort undertaken under the IAEA coordinated research project on improved instrumentation and control (I&C) maintenance techniques for research reactors. It lays the foundation for implementation of on-line monitoring (OLM) techniques and establishment of the validity of those for improved maintenance practices in research reactors for a number of applications such as change to condition based calibration, performance monitoring of process instrumentation systems, detection of process anomalies and to distinguish between process problems/effects and instrumentation/sensor issues. The techniques and guidance embodied in this publication will serve the research reactor community in providing the technical foundation for implementation of OLM techniques. It is intended to be used by Member States to implement I&C maintenance and to improve performance of research reactors.

Enhancing the functionality of reactor protection systems to provide diagnostic and monitoring information: The ISATTM approach

International Nuclear Information System (INIS)

Baldwin, J.A.; Rowe, B.J.; Jones, C.D.

1996-01-01

The ISAT TM architecture has been successfully implemented as the Single Channel Trip System (SCTS), part of the primary protection system of Nuclear Electric's Dungeness 'B' Advanced Gas-Cooled Reactors. The system is the first computer-based protection system licensed on a UK civil nuclear reactor. The system provides protection against single channel faults resulting in high coolant gas outlet temperature. The SCTS was designed to output data at several points in the system to an Ethernet to allow checks to be made on the operation of parts of the protection system and the system as a whole. In order to monitor the performance of this shutdown system a PC based monitoring system was developed to take input as data from the Ethernet, check its integrity and then analyze the data to provide information of the state of the system and subsystems. The SCTS monitor was basically intended to alert the operator to any fault on the safety system and indicate its source, provide a diagnosis of the cause of any trip initiated by the safety system, and log the occurrences of these incidents for later inspection. The intention was also to provide accurate real-time information on the thermocouple readings and to decrease the effort required to maintain the safety system. This paper will describe briefly the development of the ISAT TM monitoring system: how its requirements were arrived at, and how the design, code and testing were carried out to ensure approval for this application. It will then go on to report how the ISAT TM monitor has performed during its time in service: how more functionality has been added over and above its original requirements. Features of additional monitors for the SCTS and other ISAT TM systems will also be described. (author). 2 refs, 5 figs

Reactor inventory monitoring system for Angra-1 reactor; Sistema de monitoracao de inventario do reator para usina nuclear Angra I

Energy Technology Data Exchange (ETDEWEB)

S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M. [Furnas Centrais Eletricas S.A., Rio de Janeiro, RJ (Brazil); Soares, Milton [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Lab. de Monitoracao de Processos

1996-07-01

This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

Sodium fast reactor power monitoring using {sup 20}F tagging agent

Energy Technology Data Exchange (ETDEWEB)

Coulon, R.; Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, Centre de Saclay, 91191 Gif sur Yvette Cedex (France); Ban, G. [ENSICAEN, F-14050 Caen (France); Dumarcher, V.; Brau, H. P.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A. M.; Montagu, T.; Dautremer, T.; Barat, E.

2009-07-01

This work deals with the use of gamma spectrometry to monitor the fourth generation sodium fast reactor (SFR) power. Simulation part has shown that power monitoring in short response time and with high accuracy is possible measuring delayed gamma emitters produced in the liquid sodium. An experimental test is under preparation at French SFR Phenix experimental reactor to validate simulation studies. Physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as sodium velocity, atomic densities, neutron spectra and incident neutron cross-sections of fission reactions, and also sodium activation reactions producing gamma emitters. Then, a thermal hydraulic transfer function was used for taking into account primary sodium flow in our calculations. Gamma spectra were then determined by Monte-Carlo simulations. The experiment will be set during the reactor 'ultimate testing'. The Delayed Neutron Detection (DND) system cell has been chosen as the best available primary sodium sample for gamma power monitoring on Phenix reactor due to short sodium transit time from reactor core to measurement sample and homogenized sampling in the reactor hot pool. The main gamma spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The signal is then processed by a digital signal processing system (called Adonis) which always gives optimum answer even for high count rate and various time activity measurements. For power monitoring problematic, use of a short decay period gamma emitter as the {sup 20}F will allow to obtain a very fast response system without cumulative and flow distortion effects. These works introduce advantages and performances of this new power monitoring system for future SFR. (authors)

Utilization of local area network technology and decentralized structure for nuclear reactor core temperature monitoring

International Nuclear Information System (INIS)

Casella, M.; Peirano, F.

1986-01-01

The present system concerns Superphenix type reactors. It is a new version of system for monitoring the reactor core temperatures. It has been designed to minimize the cost and the wiring complexity because of the large number of channels (800). For this, equipments are arranged on the roof slab of the reactor with a single link to the control room; from which the name Integrated Treatment of Core Temperatures: TITC 1500 and the natural choice of a distributed system. This system monitors permanently the thermal state of the core a Superphenix type reactor. This monitoring system aims at detecting anomalies of core temperature rise, releasing automatic shutdown (safety), and providing to the monitoring systems not concerned safety the information concerning the core [fr

Reactor power distribution monitor

International Nuclear Information System (INIS)

Hoizumi, Atsushi.

1986-01-01

Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

Boiling water reactor life extension monitoring

International Nuclear Information System (INIS)

Stancavage, P.

1991-01-01

In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Real-time stability monitoring method for boiling water reactor nuclear power plants

International Nuclear Information System (INIS)

Fukunishi, K.; Suzuki, S.

1987-01-01

A method for real-time stability monitoring is developed for supervising the steady-state operation of a boiling water reactor core. The decay ratio of the reactor power fluctuation is determined by measuring only the output neutron noise. The concept of an inverse system is introduced to identify the dynamic characteristics of the reactor core. The adoption of an adaptive digital filter is useful in real-time identification. A feasibility test that used measured output noise as an indication of reactor power suggests that this method is useful in a real-time stability monitoring system. Using this method, the tedious and difficult work for modeling reactor core dynamics can be reduced. The method employs a simple algorithm that eliminates the need for stochastic computation, thus making the method suitable for real-time computation with a simple microprocessor. In addition, there is no need to disturb the reactor core during operation. Real-time stability monitoring using the proposed algorithm may allow operation under less stable margins

Full integrated system of real-time monitoring based on distributed architecture for the high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Subekti, Muhammad; Ohno, Tomio; Kudo, Kazuhiko; Takamatsu, Kuniyoshi; Nabeshima, Kunihiko

2005-01-01

A new monitoring system scheme based on distributed architecture for the High Temperature Engineering Test Reactor (HTTR) is proposed to assure consistency of the real-time process of expanded system. A distributed monitoring task on client PCs as an alternative architecture maximizes the throughput and capabilities of the system even if the monitoring tasks suffer a shortage of bandwidth. The prototype of the on-line monitoring system has been developed successfully and will be tested at the actual HTTR site. (author)

Sodium fast reactor power monitoring using gamma spectrometry

Energy Technology Data Exchange (ETDEWEB)

Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

2009-06-15

This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

The new area monitoring system and the fuel database of the TRIGA Mark II reactor in Vienna

International Nuclear Information System (INIS)

Villa, M.; Boeck, H.; Hofbauer, M.; Schwarz, V.

2004-01-01

The 250 kW TRIGA Mark-II reactor operates since March 1962 at the Atominstitut, Vienna, Austria. Its main tasks are nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection and dosimetry, and low temperature physics and fusion research. Academic research is carried out by students in the above mentioned fields coordinated and supervised by about 70 staff members with the aim of a masters- or PhD degree in one of the above mentioned areas. After 25 years of successful operation, it was necessary to exchange the old area monitoring system with a new digital one. The purpose of the new system is the permanent control of the reactor hall, the primary and secondary cooling system and the monitoring of the ventilation system. The paper describes the development and implementation of the new area monitoring system. The second topic in this paper describes the development of the new fuel database. Since March 7th, 1962, the TRIGA Mark II reactor Vienna operates with an average of 263 MWh per year, which corresponds to a uranium burn-up of 13.7 g per year. Presently we have 81 TRIGA fuel elements in the core, 55 of them are old aluminium clad elements from the initial criticality while the rest are stainless steel clad elements which had been added later to compensate the uranium consumption. Because 67 % of the elements are older than 40 years, it was necessary to put the history of every element in a database, to get an easy access to all the relevant data for every element in our facility. (author)

Incorporation of personal computers in a research reactor instrumentation system for data monitoring and analysis

International Nuclear Information System (INIS)

Leopando, L.S.

1998-01-01

The research contract was implemented by obtaining off-the shelf personal computer hardware and data acquisition cards, designing the interconnection with the instrumentation system, writing and debugging the software, and the assembling and testing the set-up. The hardware was designed to allow all variables monitored by the instrumentation system to be accessible to the computers, without requiring any major modification of the instrumentation system and without compromising reactor safety in any way. The computer hardware addition was also designed to have no effect on any existing function of the instrumentation system. The software was designed to implement only graphical display and automated logging of reactor variables. Additional functionality could be easily added in the future with software revision because all the reactor variables are already available in the computer. It would even be possible to ''close the loop'' and control the reactor through software. It was found that most of the effort in an undertaking of this sort will be in software development, but the job can be done even by non-computer specialized reactor people working with programming languages they are already familiar with. It was also found that the continuing rapid advance of personal computer technology makes it essential that such a project be undertaken with inevitability of future hardware upgrading in mind. The hardware techniques and the software developed may find applicability in other research reactors, especially those with a generic analog research reactor TRIGA console. (author)

Loose parts monitoring system

International Nuclear Information System (INIS)

Wakasa, Kohji; Nishida, Eiichi; Ishii, Kazuo; Yamanaka, Hiroto.

1987-01-01

In the loose parts monitoring system (LPMS), installed for integrity monitoring of the nuclear power plants; when there occur foreign metallic objects in the reactor primary system, including a steam generator and the piping, the sounds caused by them moving with the cooling water and thereby getting in contact with various structures are detected. Its purpose is, therefore, to detect any abnormality in the reactor plant system through such abnormal sounds due to loose or fallen supports etc., and so provide this information to the reactor operators. In principle, accelerometers are distributed in such as reactor vessel, steam generator, coolant pumps, etc., so that various sounds are collected and converted into electrical signals, followed by analysis of the data. Described are the LPMS configuration/functions, the course taken in LPMS development, future problems, etc. (Mori, K.)

Advanced in-core monitoring system for high-power reactors

International Nuclear Information System (INIS)

Mitin, V.I.; Alekseev, A.N.; Golovanov, M.N.; Zorin, A.V.; Kalinushkin, A.E.; Kovel, A.I.; Milto, N.V.; Musikhin, A.M.; Tikhonova, N.V.; Filatov, V.P.

2006-01-01

This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including WWER-1000. The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error. This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors. On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, departure from nucleate boiling ratio peaking factor) have been automatically generated. The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class IE and class B and C tasks in accordance with international standards of separation and safety category realization. Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process (Authors)

The roles of the seismic safety and monitoring systems in the PEC fast reactor

International Nuclear Information System (INIS)

Masoni, P.; Di Tullio, E.M.; Massa, B.; Martelli, A.; Sano, T.

1988-01-01

Two different seismic systems are foreseen in the case of PEC: the seismic safety system, that provides the automatic scram, and the seismic monitoring system. During earthquake, three triaxial seismic switches are triggered if a threshold value of the ground acceleration is exceeded. In this case, the signals from the seismic switches are processed by the safety system (with a 2/3 logic) and the shutdown system is triggered. Peak acceleration is the parameter used by the safety system to quantify the seismic event. This way, however, no information is obtained with regard to earthquake frequency content. Thus, reactor safety is guaranteed by adopting a threshold considerably lower than the Z.P.A. of the Design Basis Earthquake. Furthermore, in the case of significant earthquakes, the seismic motion is measured by about 20 triaxial accelerometers, located both in the free field and on the plant's structures. Data are digitazed and recordered by the seismic monitoring system. This system also elaborates the recordered time-histories providing floor response spectra and compares such spectra to the design values. The above-mentioned elaborations and comparisons are performed in short time for two triaxial measuring positions, thus allowing the Operator to immediately get a more complete information on the seismic event. The complete set of data recorded by the seismic monitoring system also allows the actual dynamic response of the plant to be determined and compared to the design values. On the basis of this comparison the necessary safety analysis can be carried out to verify whether the design limits of the plant were respected: in the positive case the reactor can be restarted. (author)

Radiation protection monitoring at the JOYO experimental fast reactor

International Nuclear Information System (INIS)

Ouchi, S.; Endo, K.; Susaki, T.

1979-01-01

This paper describes the radiation protection monitoring programme for the JOYO experimental fast reactor and some of the health physics problems experienced during the low-power nuclear tests. These include: a detailed description of the centralized radiation monitoring system; the methods and results of the individual monitoring systems; the results of operational monitoring for the handling of new plutonium fuel subassemblies; the evaluation of the external radiation dose rate around the primary coolant system; and the results of an experiment on the thermal dependence of some personnel dose meters. (author)

Operation monitoring and protection method for nuclear reactor

International Nuclear Information System (INIS)

Tochihara, Hiroshi.

1995-01-01

In an operation and monitoring method for a PWR-type reactor by using a tetra-sected neutron detector, axial off set is defined by neutron detector signals with respect to an average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core. A departure from nucleate boiling (DNBR) is represented by standardized signals, and the DNBR is calculated by using the axial off set of the average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core, and they are graphically displayed. In addition, a thermal flow rate-water channel coefficient is also graphically displayed, and the DNBR and the thermal flow rate-water channel coefficient are restricted based on the display, to determine an allowable operation range. As a result, it is possible to provide an operation monitoring and protection method for nuclear reactor capable of reducing labors and frequencies for the change of protection system setting in a case of using a tetra-sected neutron detector disposed at the outside and, at the same time, protecting each of DNR and the highest linear power or the thermal water coefficient channel. (N.H.)

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

International Nuclear Information System (INIS)

Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

2008-01-01

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

2008-01-15

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

5. symposium on nuclear reactor remote monitoring

International Nuclear Information System (INIS)

1987-01-01

17 papers deal with the data-technological concept and mode of operation of nuclear-reactor remote-monitoring (RM) systems from the perspectives of users in Baden-Wurttemberg, Sleswig-Holstein, Bavaria and Belgium, with the requirements on measuring devices and equipment in NRM systems, computer-controlled evaluation and processing of measured data, in particular the LASAT and OLDES systems. (DG) [de

Power distribution monitor in a nuclear reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1983-01-01

Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

Nuclear instrumentation systems in prototype fast breeder reactor

International Nuclear Information System (INIS)

Vijayakumaran, P.M.; Nagaraj, C.P.; Paramasivan-Pillai, C.; Ramakrishnan, R.; Sivaramakrishna, M.

2004-01-01

The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection and Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM (safety action) to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident. (authors)

Research and Development of Protection OPC server for China advanced research reactor digital monitoring system

International Nuclear Information System (INIS)

Jia Yuwen; Xu Qiguo

2012-01-01

OPC server was developed as I/O driver to communicate the digital monitoring system of China Advanced Research Reactor iFIX and protection system. The framework and working principle of the OPC server were researched, and an effective method was developed to resolve the special communication protocol. After commissioning and testing, the results show that this method is reliable and stable, makes the system easy to configure, and can reduce the complexity of the system. (authors)

Method of inspecting the function of reactor noise monitoring device

International Nuclear Information System (INIS)

Yamanaka, Hirohito.

1985-01-01

Purpose: To enable to inspect the function of a reactor noise monitoring device used for monitoring the operation abnormality in coolant circuits during reactor operation. Constitution: A cylinder incorporating a steel ball moved laterally by a pneumatic pressure is disposed to the main body of a reactor coolant circuit. A three-way solenoid valve disposed to a central control room outside to a radiation controlled area is connected with the cylinder by way of pneumatic pipeways. The three-way solenoid valve is operated for a certain period of time by a timer in the central control room to thereby impinge the steel ball in the cylinder against the main body of the coolant circuit and it is inspected as to whether the reactor noise monitoring system can detect the impinging energy or not. Accordingly, the remote control is possible from out of the radiation controlled area and the inspection work can be simplified. (Seki, T.)

Enhancing the functionality of reactor protection systems to provide diagnostic and monitoring information: The ISAT{sup TM} approach

Energy Technology Data Exchange (ETDEWEB)

Baldwin, J A; Rowe, B J [AEA Technology, Winfrith (United Kingdom); Jones, C D [Nuclear Electric Ltd., Kent (United Kingdom). Dungeness ` B` Power Station

1997-12-31

The ISAT{sup TM} architecture has been successfully implemented as the Single Channel Trip System (SCTS), part of the primary protection system of Nuclear Electric`s Dungeness `B` Advanced Gas-Cooled Reactors. The system is the first computer-based protection system licensed on a UK civil nuclear reactor. The system provides protection against single channel faults resulting in high coolant gas outlet temperature. The SCTS was designed to output data at several points in the system to an Ethernet to allow checks to be made on the operation of parts of the protection system and the system as a whole. In order to monitor the performance of this shutdown system a PC based monitoring system was developed to take input as data from the Ethernet, check its integrity and then analyze the data to provide information of the state of the system and subsystems. The SCTS monitor was basically intended to alert the operator to any fault on the safety system and indicate its source, provide a diagnosis of the cause of any trip initiated by the safety system, and log the occurrences of these incidents for later inspection. The intention was also to provide accurate real-time information on the thermocouple readings and to decrease the effort required to maintain the safety system. This paper will describe briefly the development of the ISAT{sup TM} monitoring system: how its requirements were arrived at, and how the design, code and testing were carried out to ensure approval for this application. It will then go on to report how the ISAT{sup TM} monitor has performed during its time in service: how more functionality has been added over and above its original requirements. Features of additional monitors for the SCTS and other ISAT{sup TM} systems will also be described. (author). 2 refs, 5 figs.

Leak monitoring method for a reactor container

International Nuclear Information System (INIS)

Uehara, Toshio.

1987-01-01

Purpose: To confirm leakages from a container upon nuclear reactor operation. Method: Leakages from a nuclear reactor container has been prevented by lowering the inner pressure of the container relative to the external pressure. In the conventional method of calculating the leakage by applying an inner pressure to the container and measuring the pressure change, etc. after the elapse of a pre-determined time, the measurement has to be conducted at periodical inspection when the nuclear reactor is shut-down. In view of the above, the leak test is conducted in the present invention by applying a slight inner pressure to the inside of the reactor container by supplying gases from a gas supply system and detecting the flow rate of the gases in the gas supply system while maintaining the slight inner pressure constant by controlling the supply and discharge of the gases. By applying such a inner pressure as causing no effect to the reactor operation, it is possible to monitor the leaks during operation and to detect the flow rate value surely and continuously if the leak is resulted. (Kamimura, M.)

Development of methods for monitoring and controlling power in nuclear reactors

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole dos; Silva, Vitor Vasconcelos Araujo

2012-01-01

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235 U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of

Power distribution monitoring and control in the RBMK type reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Postnikov, V.V.; Volod'ko, Yu.I.

1980-01-01

Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

International Nuclear Information System (INIS)

Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

1992-01-01

In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

Instrument failure monitoring in nuclear power systems

International Nuclear Information System (INIS)

Tylee, J.L.

1982-01-01

Methods of monitoring dynamic systems for instrument failures were developed and evaluated. In particular, application of these methods to nuclear power plant components is addressed. For a linear system, statistical tests on the innovations sequence of a Kalman filter driven by all system measurements provides a failure detection decision and identifies any failed sensor. This sequence (in an unfailed system) is zero-mean with calculable covariance; hence, any major deviation from these properties is assumed to be due to an instrument failure. Once a failure is identified, the failed instrument is replaced with an optimal estimate of the measured parameter. This failure accommodation is accomplished using optimally combined data from a bank of accommodation Kalman filters (one for each sensor), each driven by a single measurement. Using such a sensor replacement allows continued system operation under failed conditions and provides a system operator with information otherwise unavailable. To demonstrate monitor performance, a liner failure monitor was developed for the pressurizer in the Loss-of-Fluid Test (LOFT) reactor plant. LOFT is a small-scale pressurized water reactor (PWR) research facility located at the Idaho National Engineering Laboratory. A linear, third-order model of the pressurizer dynamics was developed from first principles and validated. Using data from the LOFT L6 test series, numerous actual and simulated water level, pressure, and temperature sensor failures were employed to illustrate monitor capabilities. Failure monitor design was applied to nonlinear dynamic systems by replacing all monitor linear Kalman filters with extended Kalman filters. A nonlinear failure monitor was derived for LOFT reactor instrumentation. A sixth-order reactor model, including descriptions of reactor kinetics, fuel rod heat transfer, and core coolant dynamics, was obtained and verified with test data

Monitoring-control of the 900 MWe and 1300 MWe nuclear reactors

International Nuclear Information System (INIS)

Meyer, J.

1982-01-01

After a short definition of the monitoring-control of the 900 MWe and 1300 MWe nuclear reactors, and a recall of requirements of nuclear energy, this paper presents the following points concerning the whole system of monitoring-control: the organization, the systems (instrumentation, automation), the technologies, the imperfections and the improvements brought to the system [fr

Micro processor based research reactor instrumentation and control system

International Nuclear Information System (INIS)

Hyde, W.K.

1987-01-01

The system consists of a Control System Computer (CSC) incorporated into a Reactor Control Console (RCC) and a Data Acquisition and Control Unit (DAC) adjacent to the reactor. The CSC has a high resolution color graphics CRT monitor which provides real-time graphic simulation of the reactor and a number of bar graphs displaying strategic parameters of the reactor system. In addition, abnormal or dangerous conditions are displayed. The CSC is equipped with two printers eliminating manual logging of reactor data. The reactor display and pulse mode display may also be printed. Historical data is saved in the system's large capacity memory and may be replayed and/or printed. Because of the CSC's inherent high speed math capability, raw reactor data will be quickly converted and displayed in real-time. Data can be presented in meaningful engineering units. The DAC provides a high speed data acquisition and control capability adjacent to the reactor. It continuously collects data from the reactor system, concentrates the data into a database and transmits it to the CSC when requested. Data transmission is over one of two data trunks to the CSC. The secondary trunk is used if the primary trunk fails. The data trunks drastically reduce the wiring requirements between the reactor and the Control Console. During steady-state operation of the reactor, operator commands to adjust the rod positions is transmitted from the CSC to the DAC which re-issues the commands to the drive mechanisms. In the automatic mode, the DAC will control the position of the rods via a PID algorithm. The system is independently monitored by two or more safety computers. Their function is to monitor the power level, the rate of change of power and fuel temperature of the reactor and to independently shut the reactor down in the event of a potentially dangerous (scram) condition. (author)

Development of monitoring, control and protection systems of nuclear power reactors in the USSR: Status and trends

International Nuclear Information System (INIS)

Kondrat'ev, V.V.; Mikhailov, M.N.; Prozorov, V.K.; Chudin, A.G.

1992-01-01

In 1989-90 the Supervision Body approved the new safety regulations for nuclear power plants. The operating plants do not always completely satisfy them and a great amount of work to develop operating power units up to the conditions required is necessary. This paper describes briefly the main changes made in monitoring, control and protection systems of nuclear power reactors to increase the reactor safety. The following fields are presented in the paper: General status of the NPP control and safety systems and instrumentation in USSR; sensors; electronic equipment; actuators improvement; qualification tests. (author). 3 figs

High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

International Nuclear Information System (INIS)

1983-01-01

The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

Annexes to the lecture on reactor protection system including engineered features actuation system

International Nuclear Information System (INIS)

Palmaers, W.

1982-01-01

The present paper deals with the fundamentals for a reactor protection system and discusses the following topics: - System lay-out - Analog measured data acquisition - Analog measured data processing - Limit value generation and logical gating - Procesing of the reactor protection actuation signals - Decoupling of the reactor protection system - Mechanical lay-out - Monitoring system and - Emergency control station. (orig./RW)

The application of external vibration monitoring to reactors with concrete pressure vessels

International Nuclear Information System (INIS)

Hammill, W.J.

1979-01-01

The application of external vibration monitoring techniques to advanced gas cooled reactors (AGR) which have concrete pressure vessels is considered. A monitoring system for a particular AGR coolant circuit structure is developed, whose primary objective is to detect impacting of two components, although the detection of forced vibration response is also considered. Experimental results from instrumented components in the reactor and data from rig tests on full size units have been used together with a mathematical model of some elements of the transmission path in order to establish its dynamic characteristics and relate internal component vibration to externally measured signals. The application of external vibration monitoring to the external detection of the forced vibration response of an internal reactor assembly and the remote monitoring of circulator sound output is discussed. (author)

Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor

International Nuclear Information System (INIS)

Benevenuti, Erion de Lima

2004-01-01

The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

ERMS - Environmental Radiation Monitoring System

International Nuclear Information System (INIS)

Vax, Eran; Sarusi, Benny; Sheinfeld, Mati; Levinson, Shmuel; Brandys, Irad; Sattinger, Danny; Wengrowicz, Udi; Tshuva, Avi; Tirosh, Dan

2008-01-01

A new Environmental Radiation Monitoring System (ERMS) has been developed in the NRCN as an extensive tool to be applied in case of nuclear malfunction or Nuclear Disposal Device (NDD) incident, as well as for routine radiation monitoring of the reactor's vicinity. The system collects real-time environmental data such as: gamma radiation, wind speed, wind direction, and temperature for monitoring purposes. The ERMS consists of a main Control Center and an array of monitoring stations. Fixed, environmental, gamma radiation monitoring stations are installed at the reactor's surroundings while portable stations can be posted rapidly along the wind direction, enhancing the spatial sampling of the radiation measurements and providing better hazard assessment at an emergency event. The presented ERMS, based on industrial standards for hardware and network protocols, is a reliable standalone system which upgrades the readiness to face a nuclear emergency event by supplying real-time, integrated meteorological and radiation data. (author)

Implementation of new core cooling monitoring system for light water reactors - BCCM (Becker Core Cooling Monitor)

International Nuclear Information System (INIS)

Coville, Patrick; Eliasson, Bengt; Stromqvist, Erik; Ward, Olav; Fox, Georges; Ashjian, D. T.

1998-01-01

Core cooling monitors are key instruments to protect reactors from large accidents due to loss of coolant. Sensors presented here are based on resistance thermometry. Temperature dependent resistance is powered by relatively high and constant current. Value of this resistance depends on thermal exchange with coolant and when water is no more surrounding the sensors a large increase of temperature is immediately generated. The same instrument can be operated with low current and will measure the local temperature up to 1260 o C in case of loss of coolant accident. Sensors are manufactured with very few components and materials already qualified for long term exposure to boiling or pressurized water reactors environment. Prototypes have been evaluated in a test loop up to 160 bars and in the Barsebaeck-1 reactor. Industrial sensors are now in operation in reactor Oskarshamn 2. (author)

Monitoring and control of anaerobic reactors

DEFF Research Database (Denmark)

Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

2003-01-01

The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process mea...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

Monitoring and control of anaerobic reactors

DEFF Research Database (Denmark)

Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

2003-01-01

The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

Artificial neural network for research reactor safety status monitoring

International Nuclear Information System (INIS)

Varde, P.V.

2001-01-01

During reactor upset/abnormal conditions, emphasis is placed on plant operator's ability to quickly identify the problem and perform diagnosis and initiate recovery action to ensure safety of the plant. However, the reliability of human action is adversely affected at the time of crisis, due to the time stress and psychological factors. Availability of operational aids capable of monitoring the status of the plant and quickly identifying the deviation from normal operation is expected to significantly improve the operator reliability. Artificial Neural Network (based on Back Propagation Algorithm) has been developed and applied for reactor safety status monitoring, as part of an Operator Support System. ANN has been trained for 14 different plant states using 42 input symptom patterns. Recall tests performed on the ANN show that the system was able to identify the plant state with reasonable accuracy. (author)

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor

International Nuclear Information System (INIS)

Urcia, Agustin; Arrieta, Rolando; Baltuano, Oscar; Chan, Renzo; Tincopa, Jean Pierre; Urquizo, Rafael; Rosas, Bernick

2014-01-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

Research on intelligent monitor for 3D power distribution of reactor core

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

Hardware and software concept of the remote monitoring system for nuclear reactors in Bavaria

International Nuclear Information System (INIS)

Gietl, G.

1981-01-01

The remote monitoring system for nuclear reactors (KFUe) is a fully automatic system for the measuring and registration of radioactive releases and of operation parameters of the nuclear power plants in Bavaria. It has also to measure the meteorological parameters on the site of a power plant for the purpose of dispersion calculations. The system consists of the network centre and the subsystems with their satellite stations at the nuclear power plants. Process computers in the network centre and in the subsystems perform the automatic operation and they handle the data transmission via data lines. The hardware of the entire system is so conceived that the mesured data can be processed, stored and displayed in an optimal way for the user. The software of the KFUe centre and the subsystems is structured strictly modular. Redunancies are implied in the hardware and software. The automatic remote air quality monitoring system of Bavaria (LUeB) can be operated as a stand-by system for the KFUe and vice versa in case of an emergency. (orig./HP) [de

Construction of the monitoring, processing and logging systems supporting for management, operation and maintenance of the Dalat reactor control system

International Nuclear Information System (INIS)

Trinh Dinh Hai; Nguyen Thanh Cuong; Huynh Ton Nghiem; Phan Quoc Minh; Nguyen Duc Tuan; Nguyen Nhi Dien

2004-01-01

From 1/2002 to 12/2003, we implemented successfully a project, entitled 'Construction of the monitoring, processing and logging systems supporting for management, operation and maintenance of the Dalat reactor control system' under the assistance of the Ministry of Science and Technology. Its main results such as Testing Apparatus based on microcontroller for all functional boards of the Control Logic System of the Reactor Control System (RCS). Technical support CD - ROM for Process Instrumentation System, software for logging automatically information from important systems of the RCS through LAN, program for failure management of Process Instrumentation System have been playing an important role for observation, operation support, maintenance of the RCS. Through this project, the implementation group has grown up rapidly. The control and instrumentation group has been provided with some modern equipment, electronic components, and materials for maintenance work and research development in the years to come. This paper presents typical results and discussions. (author)

An on-line adaptive core monitoring system

International Nuclear Information System (INIS)

Verspeek, J.A.; Bruggink, J.C.; Karuza, J.

1997-01-01

An on-line core monitoring system has been in operation for three years in the Dodewaard Nuclear Power Plant. The core monitor uses the on-line measured reactor data as an input for a power distribution calculation. The measurements are frequently performed. The system is used for monitoring as well as for predicting purposes. The limiting thermal hydraulic parameters are monitored as well as the pellet-clad interaction limits. The data are added to a history file used for cycle burn-up calculations and trending of parameters. The reactor states are presented through a convenient graphical user interface. (authors)

Design of Reactor Coolant Pump Seal Online Monitoring System

International Nuclear Information System (INIS)

Ah, Sang Ha; Chang, Soon Heung; Lee, Song Kyu

2008-01-01

As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation

Design of Reactor Coolant Pump Seal Online Monitoring System

Energy Technology Data Exchange (ETDEWEB)

Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

2008-05-15

As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

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

Reactor cover gas monitoring at the Fast Flux Test Facility

Energy Technology Data Exchange (ETDEWEB)

Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

1987-07-01

The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

International Nuclear Information System (INIS)

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Vermeeren, L.; Lopez, A. Legrand

2011-01-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10 20 n/cm 2 . A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Science.gov (United States)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Energy Technology Data Exchange (ETDEWEB)

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

2011-03-15

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

Design of Safety Parameter Monitoring Function in a Research Reactor Facility

Energy Technology Data Exchange (ETDEWEB)

Park, Jaekwan; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

Design of Safety Parameter Monitoring Function in a Research Reactor Facility

International Nuclear Information System (INIS)

Park, Jaekwan; Suh, Yongsuk

2014-01-01

The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

Multiple microprocessor based nuclear reactor power monitor

International Nuclear Information System (INIS)

Lewis, P.S.; Ethridge, C.D.

1979-01-01

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

Water level monitoring device in nuclear reactor

International Nuclear Information System (INIS)

Miura, Kiyohide; Otake, Tomohiro.

1988-01-01

Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, J.; Vocka, R.

2010-01-01

In this work authors present 12 years of operation experience of core monitoring and surveillance system with advanced capabilities on nuclear power plants on 6 unit of VVER-440 type of reactors at two different NPPs. The original version of the SCORPIO (Surveillance of reactor CORe by PIcture On-line display) system was developed for the western type of PWR reactors. The first version of the SCORPIO-VVER Core Monitoring System for Dukovany NPP (VVER-440 type of reactor, Czech Republic) was developed in 1998. For SCORPIO-VVER implementation at Bohunice NPP in Slovakia (2001) the system was enhanced with startup module KRITEX.

Principle of human system interface (HSI) design for new reactor console of PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Mohd Sabri Minhat; Izhar Abu Hussin

2013-01-01

Full-text: This paper will describe the principle of human system interface design for new reactor console in control room at TRIGA reactor facility. In order to support these human system interface challenges in digital reactor console. Software-based instrumentation and control (I and C) system for new reactor console could lead to new human machine integration. The proposed of Human System Interface (HSI) which included the large display panels which shows reactor status, compact and computer-based workstations for monitoring, control and protection function. The proposed Human System Interface (HIS) has been evaluated using various human factor engineering. It can be concluded that the Human System Interface (HIS) is designed as to address the safety related computer controlled system. (author)

FFTF fission gas monitor computer system

International Nuclear Information System (INIS)

Hubbard, J.A.

1987-01-01

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents control room alarms and displays, and records and prints data and analysis reports. The fission gas monitor system makes extensive use of commercially available hardware and software, providing a reliable and easily maintained system. The design provides extensive automation of previous manual operations, reducing the need for operator training and minimizing the potential for operator error. The dual nature of the system allows one monitor to be taken out of service for periodic tests or maintenance without interrupting the overall system functions. A built-in calibrated gamma source can be controlled by the computer, allowing the system to provide rapid system self tests and operational performance reports

Nuclear reactor trip system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out 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. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

Design features of HANARO Neutron Flux Monitoring System and its operating experiences

International Nuclear Information System (INIS)

Kim, Young-Ki; Ahn, Guk-Hoon

1999-01-01

The Neutron Flux Monitoring System for HANARO provides reliable neutron flux measurement from reactor shutdown to reactor full power level ranging 10 decades from 10 0 nv to 10 10 nv. The neutron flux monitoring system consists of a guarded fission chamber, amplifier and signal processor. The neutron flux as the measure of reactor power is continuously monitored by six(6) fission chambers mounted on the courtside wall of the reflector tank in the pool. Three(3) of the fission chambers are used for reactor power control, while the other three(3) are used for tripping the reactor in case of power excursion. Only the wide range fission chamber-based neutron monitoring system is employed for neutron power measurement thereby source range and intermediate range detectors are not necessary and the number of neutron monitoring channels are minimized at HANARO. (author)

Fuel failure monitoring system design approach for KALIMER

International Nuclear Information System (INIS)

Song, Soon Ja; Hwang, I. K.; Kwon, Kee Choon

1998-01-01

Fuel Failure Monitoring System (FFMS) detects fission gas and locates failed fuels in Liquid Metal Reactor. This system comprises three subsystems; delayed neutron monitoring, cover gas monitoring, and gas tagging. The purpose of this system is to improve the integrity and availability of the liquid metal plant. In this paper, FFMS was analyzed on detection method and compared with various existing liquid metal plants. Sampling and detecting methods were classified with specific plant types. Several technologies of them was recognized and used in most liquid metal reactors. Detection technology and analysis performance, however, must be improved because of new technology when liquid metal plant is built, but the FFMS design scheme will not be changed. Thereby this paper suggests the design to implement KALIMER(Korea Advanced LIquid MEtal Reactor) FFMS

Design considerations for post accident monitoring system of a research reactor

International Nuclear Information System (INIS)

Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki

2012-01-01

The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR

Radioactivity, radiation protection and monitoring during dismantling of light-water reactors

International Nuclear Information System (INIS)

Hummel, L.; Zech, J.B.

2005-01-01

Based on the radioactivity inventory in the systems and components of light-water reactors observed during operation, the impact of actions during plant emptying after the conclusion of power operation and possible subsequent long-term safe enclosure concerning the composition of the nuclide inventory of the plant to be dismantled will be described. Derived from this will be the effects on radioactivity monitoring in the plant, physical radiation protection monitoring, and the measured characterization of the residual materials resulting from the dismantling. The impact of long-term interim storage will also be addressed in the discussion. The talk should provide an overview of the interrelationships between source terms, decay times and the radioactivity monitoring requirements of the various dismantling concepts for commercial light-water reactors. (orig.)

Reactor Emergency Action Level Monitor

International Nuclear Information System (INIS)

Touchton, R.A.

1987-01-01

The Reactor Emergency Action Level Monitor (REALM) Expert System is designed to provide assistance in the identification of a nuclear power plant emergency situation and the determination of its severity. REALM has been developed to operate in a real-time processing environment. REALM embodies a hybrid architecture utilizing both rule-based reasoning and object-oriented programming techniques borrowed from the Artificial Intelligence discipline of Computer Sciences. The rulebase consists of event-based rules and symptom-based rules. The symptom-based rules go beyond the current EAL structure to address the more problematic scenarios and entail a more symbolic representation of the plant information. The results to date have been encouraging that expert system technology can provide improved emergency decision-making capability in nuclear power plants

Development of a risk monitor for research reactor

International Nuclear Information System (INIS)

Agrawal, Mayank; Mathur, D.; Joshi, N.S.; Varde, P.V.

2014-01-01

The traditional approach to engineering asset management, which primarily includes management and operation and maintenance activities, in a nuclear plant is based on engineering judgment. Here, the defence-in-depth and conservative criteria form the basis of decisions. Ideally, it is expected that the decisions should be based on performance of systems, structures and components (SSC), which should also take plant safety margin into consideration. Probabilistic methods are widely used as a compliment to deterministic methods in assessing safety of nuclear reactors. A risk-based operations and maintenance management system has been developed for a research reactor. The results of a level-1 PSA study in terms of minimal cut-set equations along with other engineering plant information forms major input for this system. The system is expected to be operational in plant control room conditions in support of routine decisions. This paper presents one of the main applications of the plant Level-1 Probabilistic Safety Assessment, developed from plant specific data i.e. a risk monitor. The risk monitor has several applications that support in decision making. It is expected that the use of this system in real-time scenario will provide improved understanding of safety on one hand while enabling enhancement of plant economics on the other. Some the case studies are discussed here. (author)

Radiological monitoring related to the operation of PUSPATI's Triga Reactor

International Nuclear Information System (INIS)

Fatimah Mohamad Amin; Mohamad Yusof Mohamad Ali; Lau How Mooi; Idris Besar.

1983-01-01

Reactor operation is one of the main activities carried out at the Tun Ismail Atomic Research Centre (PUSPATI) which requires radiological monitoring. This paper describes the programme for radiological monitoring which is related to the operation of the 1 MW Triga MK II research reactor which was commissioned in July, 1982. This programme includes monitoring of the radiation and contamination levels of the reactor and its associated facilities and environmental monitoring of PUSPATI's site and its environs. The data presented in this paper covers the period between 1982 to 1983 which includes both the pre-operational and operational phases of the monitoring programme. (author)

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

Reactor power distribution monitor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1980-01-01

Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

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

Monitoring and reviewing research reactor safety in Australia

International Nuclear Information System (INIS)

Cairns, R.C.; Greenslade, G.K.

1990-01-01

Th research reactors operated by the Australian Nuclear Science and Technology Organization (ANSTO) comprise the 10 MW reactor HIFAR and the 100 kW reactor Moata. Although there are no power reactors in Australia the problems and issues of public concern which arise in the operation of research reactors are similar to those of power reactors although on a smaller scale. The need for independent safety surveillance has been recognized by the Australian Government and the ANSTO Act, 1987, required the Board of ANSTO to establish a Nuclear Safety Bureau (NSB) with responsibility to the Minister for monitoring and reviewing the safety of nuclear plant operated by ANSTO. The Executive Director of ANSTO operates HIFAR subject to compliance with requirements and arrangements contained in a formal Authorization from the Board of ANSTO. A Ministerial Direction to the Board of ANSTO requires the NSB to report to him, on a quarterly basis, matters relating to its functions of monitoring and reviewing the safety of ANSTO's nuclear plant. Experience has shown that the Authorization provides a suitable framework for the operational requirements and arrangements to be organised in a disciplined and effective manner, and also provides a basis for audits by the NSB by which compliance with the Board's safety requirements are monitored. Examples of the way in which the NSB undertakes its monitoring and reviewing role are given. Moata, which has a much lower operating power level and fission product inventory than HIFAR, has not been subject to a formal Authorization to date but one is under preparation

Reactor operation monitor

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1982-01-01

Purpose: To improve the working performance of a reactor by extending the range for the power conditioning due to the control rod operation and flow rate control. Constitution: The results of calculations for the power distribution and the burn-up degree distribution of the reactor core from a reactor performance computer that processes each of measuring signals in a nuclear power plant are used as the inputs for a computing device of the fuel rod power hysteresis to form the power hysteresis for each of the fuel rods up to the present time. The data are used as the inputs for the computing device of the fuel rod performance index, and the fuel rod performance index representing the critical values for the stresses in the fuel rod cladding tubes and the critical values for the duration of the stresses determined from the power hysteresis and the burn-up degree of the fuel rod are calculated for each of the fuel rods. Accordingly, the power conditioning can be carried out upon power-up in the reactor while monitoring the fuel rod performance index f(t) for each of the fuel assemblies, whereby the range for the power conditioning due to the control rod operation and the flow rate control can be extended relative to fuel assemblies in which f(t) is smaller than 1. (Yoshino, Y.)

Neutron dosimetry. Environmental monitoring in a BWR type reactor

International Nuclear Information System (INIS)

Tavera D, L.; Camacho L, M.E.

1991-01-01

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

Summary of detection, location, and characterization capabilities of AE for continuous monitoring of cracks in reactors

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Friesel, M.A.; Pappas, R.A.; Skorpik, J.R.; Dawson, J.F.

1984-10-01

The objective of the program is to develop acoustic emission (AE) methods for continuous monitoring of reactor pressure boundaries to detect and evaluate crack growth. The approach involves three phases: develop relationships to identify crack growth AE signals and to use identified crack growth AE data to estimate flaw severity; evaluate and refine AE/flaw relationships through fatigue testing a heavy section vessel under simulated reactor conditions; and demonstrate continuous AE monitoring on a nuclear power reactor system

Tools and techniques for ageing predictions in nuclear reactors through condition monitoring

International Nuclear Information System (INIS)

Verma, R.M.P.

1994-01-01

To operate the nuclear reactors beyond their design predicted life is gaining importance because of huge replacement and decommissioning costs. But experience shows that nuclear plant safety and reliability may decline in the later years of plant life due to ageing degradation. Ageing of nuclear plant components, structures and systems, if unmitigated reduces their safety margins provided in the design and thus increases risks to public health and safety. These safety margins must be monitored throughout plant service life including any extended life. Condition monitoring of nuclear reactor components/equipment and systems can be done to study the effect of ageing, status of safety margins and effect of corrective and mitigating actions taken. The tools and techniques of condition monitoring are also important in failure trending, predictive maintenance, evaluation of scheduled maintenance, in mitigation of ageing, life extension and reliability studies. (author). 1 fig., 1 annexure

Power monitoring in space nuclear reactors using silicon carbide radiation detectors

Science.gov (United States)

Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

2005-01-01

Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

Real Time Corrosion Monitoring in Lead and Lead-Bismuth Systems

Energy Technology Data Exchange (ETDEWEB)

James F. Stubbins; Alan Bolind; Ziang Chen

2010-02-25

The objective of this research program is to develop a real-time, in situ corrosion monitoring technique for flowing liquid Pb and eutectic PbBi (LBE) systems in a temperature range of 400 to 650 C. These conditions are relevant to future liquid metal cooled fast reactor operating parameters. THis program was aligned with the Gen IV Reactor initiative to develp technologies to support the design and opertion of a Pb or LBE-cooled fast reactor. The ability to monitor corrosion for protection of structural components is a high priority issue for the safe and prolonged operation of advanced liquid metal fast reactor systems. In those systems, protective oxide layers are intentionally formed and maintained to limit corrosion rates during operation. This program developed a real time, in situ corrosion monitoring tecnique using impedance spectroscopy (IS) technology.

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)

Artificial neural networks versus conventional methods for boiling water reactor stability monitoring

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der

1995-01-01

The application of an artificial neural network (ANN) for boiling water reactor (BWR) stability monitoring was studied. A three-layer perceptron was trained on synthetic autocorrelation functions to estimate the decay ratio and the resonance frequency from measured neutron noise. Training of the ANN was improved by adding noise to the training patterns and by applying nonconventional error definitions in the generalized delta rule. The performance of the developed ANN was compared with those of conventional stability monitoring techniques. Explicit care was taken for generating unbiased test data. It is found that the trained ANN is capable of monitoring the stability of the Dodewaard BWR for four specific cases. By comparing properties such as the false alarm ratio, the alarm failure ratio, and the average time to alarm, it is shown that it performs worse than model-based methods in stability monitoring of exact second-order systems but that it is more robust (better resistant to corruptions of the input data and to deviations of the system at issue from an exact second-order system) than other methods. The latter explains its good performance on the Dodewaard BWR and is promising for the application of an ANN for stability monitoring of other reactors and for other operating conditions

Reactivity Monitoring of Accelerator-Driven Nuclear Reactor Systems

NARCIS (Netherlands)

Uyttenhove, W.

2016-01-01

This thesis provides a methodology and set-up of a reactivity monitoring tool for Accelerator-Driven Systems (ADS). The reactivity monitoring tool should guarantee the operation of an ADS at a safe margin from criticality. Robustness is assured in different aspects of the monitoring tool: the choice

Axial power monitoring uncertainty in the Savannah River Reactors

International Nuclear Information System (INIS)

Losey, D.C.; Revolinski, S.M.

1990-01-01

The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

Programming of computers for the protection system for Savannah River reactors

International Nuclear Information System (INIS)

Finley, R.H.

1977-06-01

The monitoring requirements for the SRP Safety Computers are shown. These fast response times coupled with the large number of analog inputs to be scanned imposed stringent program requirements. The system consists of two separate computers, each with its own inputs to monitor half the reactor positions. Either computer can provide the minimum required monitoring. The desired redundant monitoring is provided when both computers are on-line. If both computers are off-line, the reactor is automatically shut down

Development of an advanced antineutrino detector for reactor monitoring

Energy Technology Data Exchange (ETDEWEB)

Classen, T., E-mail: classen2@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bernstein, A.; Bowden, N.S. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Cabrera-Palmer, B. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Ho, A.; Jonkmans, G. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada); Kogler, L.; Reyna, D. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Sur, B. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada)

2015-01-21

Here we present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. This paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass per detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.

Preliminary Development of Online Monitoring Acoustic Emission System for the Integrity of Research Reactor Components

Science.gov (United States)

Bakhri, S.; Sumarno, E.; Himawan, R.; Akbar, T. Y.; Subekti, M.; Sunaryo, G. R.

2018-02-01

Three research reactors owned by BATAN have been more than 25 years. Aging of (Structure, System and Component) SSC which is mainly related to mechanical causes become the most important issue for the sustainability and safety operation. Acoustic Emission (AE) is one of the appropriate and recommended methods by the IAEA for inspection as well as at the same time for the monitoring of mechanical SSC related. However, the advantages of AE method in detecting the acoustic emission both for the inspection and the online monitoring require a relatively complex measurement system including hardware software system for the signal detection and analysis purposes. Therefore, aim of this work was to develop an AE system based on an embedded system which capable for doing both the online monitoring and inspection of the research reactor’s integrity structure. An embedded system was selected due to the possibility to install the equipment on the field in extreme environmental condition with capability to store, analyses, and send the required information for further maintenance and operation. The research was done by designing the embedded system based on the Field Programmable Gate Array (FPGA) platform, because of their execution speed and system reconfigurable opportunities. The AE embedded system is then tested to identify the AE source location and AE characteristic under tensile material testing. The developed system successfully acquire the AE elastic waveform and determine the parameter-based analysis such as the amplitude, peak, duration, rise time, counts and the average frequency both for the source location test and the tensile test.

Computer monitoring system for pilot plant nuclear criticality facility (solution ZPR)

International Nuclear Information System (INIS)

Hua Xiaokang; Liang Huiping

1999-01-01

The system is used for the Solution Zero Power Reactor physics measurement and safety monitoring. Its software modularization design enables multi-task real-time monitoring and off-time data processing. The system is labor/time saving to experimenters and will enhance the experiment precision and the reactor operation safety performance

Development of a nuclear reactor control system simulator using virtual instruments

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

2011-01-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

Development of a nuclear reactor control system simulator using virtual instruments

Energy Technology Data Exchange (ETDEWEB)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

Reactor monitor

International Nuclear Information System (INIS)

Takada, Tamotsu.

1992-01-01

The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

Longtime radionuclide monitoring in the vicinity of Salaspils nuclear reactor

International Nuclear Information System (INIS)

Riekstina, D.; Berzins, J.; Krasta, T.; Skrypnik, O.; Alksnis, J.

2016-01-01

The research nuclear reactor in Salaspils was decommissioned in 1998. Now reactor is partially dismantled and its territory is used as a temporary storage of radioactivity contaminated materials and water. Environment radioactivity monitoring for presence of artificial radionuclides in the vicinity of Salaspils nuclear reactor is carried out since 1990. Data include Cs-137 concentration in soils, tritium concentration in ground water, as well as H-3, Cs-137, Co-60 concentration and gross beta-activity of reactors sewage and rainwater drainage. The systematic monitoring allowed to detect in December 2014 a leakage from the special wastewater basin and so to prevent a pollution of ground water outside reactors territory.

A study of reactor monitoring method with neural network

Energy Technology Data Exchange (ETDEWEB)

Nabeshima, Kunihiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

The purpose of this study is to investigate the methodology of Nuclear Power Plant (NPP) monitoring with neural networks, which create the plant models by the learning of the past normal operation patterns. The concept of this method is to detect the symptom of small anomalies by monitoring the deviations between the process signals measured from an actual plant and corresponding output signals from the neural network model, which might not be equal if the abnormal operational patterns are presented to the input of the neural network. Auto-associative network, which has same output as inputs, can detect an kind of anomaly condition by using normal operation data only. The monitoring tests of the feedforward neural network with adaptive learning were performed using the PWR plant simulator by which many kinds of anomaly conditions can be easily simulated. The adaptively trained feedforward network could follow the actual plant dynamics and the changes of plant condition, and then find most of the anomalies much earlier than the conventional alarm system during steady state and transient operations. Then the off-line and on-line test results during one year operation at the actual NPP (PWR) showed that the neural network could detect several small anomalies which the operators or the conventional alarm system didn't noticed. Furthermore, the sensitivity analysis suggests that the plant models by neural networks are appropriate. Finally, the simulation results show that the recurrent neural network with feedback connections could successfully model the slow behavior of the reactor dynamics without adaptive learning. Therefore, the recurrent neural network with adaptive learning will be the best choice for the actual reactor monitoring system. (author)

Progress toward acoustic emission characterization for continuous monitoring of reactor pressure vessels

International Nuclear Information System (INIS)

Hutton, P.H.; Schwenk, E.B.; Kurtz, R.J.

1980-01-01

The purpose of this work is to evaluate the feasibility of detecting and analyzing flaw growth in an operating reactor vessel using acoustic emission (AE) data. A preliminary AE-flaw growth relationship has been developed encompassing six variables. This model relates AE count and energy to fatigue crack driving force ΔK. Pattern recognition has been examined as a means of distinguishing crack growth AE from other acoustic signals with encouraging results. The program is moving into hardware implementation of relationships for testing and refinement by monitoring tests in heavy section specimens and subscale structures. The hardware system will become a reactor monitor prototype. (orig.) [de

Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

International Nuclear Information System (INIS)

Richard P. Wells

2007-01-01

This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

Reactor power monitoring device

International Nuclear Information System (INIS)

Kono, Shigehiro.

1990-01-01

Among a plurality of power monitoring programs in a reactor power monitoring device, rapid response is required for a scram judging program for the power judging processing of scram signals. Therefore, the scram judging program is stored independently from other power monitoring programs, applied with a priority order, and executed in parallel with other programs, to output scram signals when the detected data exceeds a predetermined value. As a result, the capacity required for the scram judging program is reduced and the processing can be conducted in a short period of time. In addition, since high priority is applied to the scram judging program which is divided into a small capacity, it is executed at higher frequency than other programs when they are executed in parallel. That is, since the entire processings for the power monitoring program are repeated in a short cycle, the response speed of the scram signals required for high responsivity can be increased. (N.H.)

Tritium monitoring within the reactor hall of a DT fusion reactor

International Nuclear Information System (INIS)

Jalbert, R.A.

1983-01-01

Monitoring the reactor hall atmosphere of DT-fueled fusion reactors will probably be performed with conventional ion chamber and proportional counter instruments modified as necessry to deal with the background radiation. Background includes external neutron and gamma radiation and internal beta-gamma radiation from the activated atmosphere. Although locating instruments in remote areas of the reactor hall and adding local shielding and electronic compensation may be feasible, placing the instruments in accessible low-background areas outside of the reactor hall and doing remote sampling is preferable and solves most of the radiation problems. The remaining problem of the activated atmosphere may be solved by recently developed instruments in conjunction with the use of semi-permeable membranes currently under development and evaluation

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)

A computer control system for a research reactor

International Nuclear Information System (INIS)

Crawford, K.C.; Sandquist, G.M.

1987-01-01

Most reactor applications until now, have not required computer control of core output. Commercial reactors are generally operated at a constant power output to provide baseline power. However, if commercial reactor cores are to become load following over a wide range, then centralized digital computer control is required to make the entire facility respond as a single unit to continual changes in power demand. Navy and research reactors are much smaller and simpler and are operated at constant power levels as required, without concern for the number of operators required to operate the facility. For navy reactors, centralized digital computer control may provide space savings and reduced personnel requirements. Computer control offers research reactors versatility to efficiently change a system to develop new ideas. The operation of any reactor facility would be enhanced by a controller that does not panic and is continually monitoring all facility parameters. Eventually very sophisticated computer control systems may be developed which will sense operational problems, diagnose the problem, and depending on the severity of the problem, immediately activate safety systems or consult with operators before taking action

Wireless condition monitoring for the RA-6 research reactor

International Nuclear Information System (INIS)

Garcia Peyrano, O.; Calzeta, O.; Rico, N.; Damiani, H.; Coutsiers, E.

1999-01-01

The vibration laboratory at C.A.B. has a great experience with the analysis and diagnostic of symptoms of failures in the rotating equipment of the R-6 research reactor and in our longest NPP (CANDU 600 Mw), located in Embalse town, Cordoba City, Argentina. Objective: The standard condition monitoring instrumentation system were designed for large equipment operating under different environmental conditions and sensitivities. The signal processing is not flexible and the diagnostic is an expensive method for the small poll type research reactors. This papers describes the research and development which are related whit the new concept, cheaper and flexible condition monitoring instrumentation system. Implementing a vibration analysis measurements technique with a sensor inside (in the pool) of the nuclear reactor RA-6, and mainly based on fft signal processing, an extensive program for vibration source identification was done. Different nuclear power conditions were monitored as full power and in zero power, also. This zero power shows the best acoustical environmental, because the cooling pumps are stop, and the core is cooling by natural convection. Two sensors were mainly used as the detector's subsystem. One of these detectors was an accelerometer attached to the top of the fine control rod and the other one was a water resistant omnidirectional microphone which was located underwater at different distances from the nuclear core. All the signal measurement by this two sensors were recorded and then was processed. Both signal was acquired at the same time for correlation analysis purposes. The analysis was composed by a 'Spectral Dynamics SD380' connected to a P.C. with dedicated post processing software. On the other hand, some calibration and sensitivity comparison was done using an SKFCM40, dual channel data collector and analyzer. (author)

BEACON TSM application system to the operation of PWR reactors

International Nuclear Information System (INIS)

Lozano, J. A.; Mildrum, C.; Serrano, J. F.

2012-01-01

BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

Reactor protection systems of 500 MWe PHWRs

Energy Technology Data Exchange (ETDEWEB)

Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

1997-03-01

The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

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.)

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

International Nuclear Information System (INIS)

Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

2014-01-01

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

HUMOS monitoring system of leaks into the containment atmosphere

International Nuclear Information System (INIS)

Matal, O.; Zaloudek, J.; Matal, O. Jr.; Klinga, J.; Brom, J.

1997-01-01

The detection and monitoring of coolant leaks into the containment atmosphere during reactor operation is a major safety measure. Using the HUMOS monitoring system, leaks can be detected in pressure tests of integrity and in any other mode of operation when the reactor ventilation system is operating and the primary circuit and its components are pressurized. Performance tests, the design, hardware and software of the HUMOS system are briefly described. A test was performed to demonstrate that a small amount of humidity released by leakage into the containment air can be detected. (M.D.)

Continuous Monitoring of GAMMA Radiation Field in the Reactor RA Building

International Nuclear Information System (INIS)

Stalevski, T.

2008-01-01

This paper presents the system for continuos monitoring of gamma doze rate in the reactor RA building. Industrial (PC compatible) computer acquires analog signals from eight ionization chambers and eight analog signals from three BPH devices. Digital output interface is used for testing ionization chambers and BPH devices. Computer program for data analyzes and presentation is written in graphical programming language LabVIEW and enables monitoring of measured data in real time. Measured data can be monitored over local computer network, Internet and mobile devices using standard web browsers. (author)

observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

International Nuclear Information System (INIS)

Siry, S.A K.

2007-01-01

analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

Online NPP monitoring with neuro-expert system

International Nuclear Information System (INIS)

Nabeshima, K.

2002-01-01

This study present a hybrid monitoring system for nuclear power plant utilizing neural networks and a rule-based expert system. The whole monitoring system including a data acquisition system and the advisory displays has been tested by an on-line simulator of pressurized water reactor. From the testing results, it was shown that the neural network in the monitoring system successfully modeled the plant dynamics and detected the symptoms of anomalies earlier than the conventional alarm system. The expert system also worked satisfactorily in diagnosing and displaying the system status by using the outputs of neural networks and a priori knowledge base

Data acquisition, monitoring and diagnostic system for predictive control and protection of rotating components of IEAR-1 reactor by vibration analysis

International Nuclear Information System (INIS)

Serra, Reynaldo Cavalcanti; Tecco, Dorival Goncalves

1996-01-01

This work presents the vibration and temperature data acquisition, monitoring and diagnostic systems, recently installed in the primary circuit, secondary circuit and emergency generator of the IEA-R1 reactor at IPEN during the course of the first power elevation tests to 5MW. It incorporates a series of routines for equipment configuration, interactive automatic monitoring , data processing and documentation/storage without the exposure of operators in the radiological protection areas. (author)

Research of management information system of radiation protection for low temperature nuclear heating reactor

International Nuclear Information System (INIS)

Bai Hongtao; Wang Jiaying; Wu Manxue

2001-01-01

Management information system of radiation protection for low temperature reactor uses computer to manage the data of the low temperature nuclear heating reactor radiation monitoring, it saves the data from the front real-time radiation monitoring system, comparing these data with historical data to give the consequence. Also, the system provides some picture in order to show space information at need. The system, based on Microsoft Access 97, consists of nine parts, including radiation dose, environmental data, meteorological data and so on. The system will have value in safely operation of the low temperature nuclear heating reactor

Facility Effluent Monitoring Plan for the N Reactor

International Nuclear Information System (INIS)

Watson, D.J.; Brendel, D.F.; Shields, K.D.

1991-11-01

A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The primary purpose of the N Reactor Facility Effluent Monitoring Plan (FEMP), during standby, is to ensure that the radioactive effluents are properly monitored and evaluated for compliance with the applicable DOE orders and regulatory agencies at the federal, state, and local levels. A secondary purpose of the FEMP is to ensure that hazardous wastes are not released, in liquid effluents, to the environment even though the potential to do so is extremely low. The FEMP is to provide a monitoring system that collects representative samples in accordance with industry standards, performs analyses within stringent quality control (QC) requirements, and evaluates the data through the use of comparative analysis with the standards and acceptable environmental models

ZPR-9 airborne plutonium monitoring system

International Nuclear Information System (INIS)

Rusch, G.K.; McDowell, W.P.; Knapp, W.G.

1975-01-01

An airborne plutonium monitoring system which is installed in the ZPR-9 (Zero Power Reactor No. 9) facility at Argonne National Laboratory is described. The design and operational experience are discussed. This monitoring system utilizes particle size and density discrimination, alpha particle energy discrimination, and a background-subtraction techique operating in cascade to separate airborne-plutonium activity from other, naturally occurring, airborne activity. Relatively high sensitivity and reliability are achieved

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)

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

Science.gov (United States)

Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

2015-12-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

International Nuclear Information System (INIS)

Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

2015-01-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

Energy Technology Data Exchange (ETDEWEB)

Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

2015-12-15

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

Environmental radiation monitoring around the research reactors

International Nuclear Information System (INIS)

Lee, Chang Woo; Lee, Hyun Duk; Kim, Sam Rang; Choi, Yong Ho; Kim, Jeong Moo; Lee, Myeon Joo; Lee, Myeong Ho; Hong, Kwang Hee; Lim, Moon Ho; Lee, Won Yoon; Park, Do Won; Choi, Sang Do

1993-04-01

Environmental radiation monitoring was carried out measurement of environmental radiation and environmental radioactivity analysis around KAERI nuclear facilities and Seoul research reactor. The results of environmental radiation monitoring around KAERI nuclear facilities and Seoul research reactor are the follows : The average level of environmental radiation dose measured by NaI scintillation counter and accumulated radiation dose by TLD was almost same level compared with the past years. Gross α, β radioactivity in environmental sample was not found abnormal data. γ-radionuclides in water samples were not detected. But only radionuclide K-40, which is natural radionuclide, was detected in the all samples and Cs-137 was detected in the surface soil and discharge sediment. The average level of environmental radiation dose around Seoul research reactor was almost same level compared with the past years, and Be-7 and Cs-137 were detected in some surface soil and discharge sediment by γ-spectrometry. (Author)

Diagnostic and monitoring systems in nuclear power plants

International Nuclear Information System (INIS)

Wehling, H.J.; Jax, P.; Streicher, V.

1987-01-01

Monitoring systems are important for the availability of nuclear power plants. A survey is given about such systems designed and constructed by the Kraftwerk Union AG Erlangen (Federal Republic of Germany) in order to assure the mechanical integrity of reactor cooling systems. Three monitoring systems based on microprocessors are presented: KUES (acoustic detection of loose parts), SUES (vibration), and FAMOS (fatigue)

The JOYO remote monitoring system

International Nuclear Information System (INIS)

Damico, Joseph P.; Hashimoto, Yu

2000-01-01

The evolution of the personal computer, operating systems and applications software and the Internet has brought drastic change and many benefits worldwide. Remote monitoring systems benefit from computer network and other modern software technologies. The availability of fast, inexpensive and secure communications enables new solutions for monitoring system applications. The JOYO Remote Monitoring System (RMS) utilizes computer network communications and modular software design to provide a distributed integrated solution for monitoring multiple storage locations. This paper describes the remote monitoring system installed at the JOYO Fast Reactor. The system combines sensors, software, and computer network technologies to create a powerful data collection, storage and dissemination capability. The RMS provides a flexible, scalable solution for a variety of applications. The RMS integrates a variety of state of the art technologies from several sources and serves as a test bed for cutting edge technologies that can be shared with outside users. This paper describes the system components and their operation and discusses system benefits. Current activities and future plants for the JOYO RMS will be discussed. (author)

Temperature monitoring of gas-cooled reactors

International Nuclear Information System (INIS)

Kaiser, G.E.

1977-01-01

The present paper deals with questions like : a) Why temperature monitoring in high-temperature reactors at all. b) How are the measuring positions arranged and how are the measurements designed. c) What technique of temperature measurement is applied. (RW) [de

A Best-Estimate Reactor Core Monitor Using State Feedback Strategies to Reduce Uncertainties

International Nuclear Information System (INIS)

Martin, Robert P.; Edwards, Robert M.

2000-01-01

The development and demonstration of a new algorithm to reduce modeling and state-estimation uncertainty in best-estimate simulation codes has been investigated. Demonstration is given by way of a prototype reactor core monitor. The architecture of this monitor integrates a control-theory-based, distributed-parameter estimation technique into a production-grade best-estimate simulation code. The Kalman Filter-Sequential Least-Squares (KFSLS) parameter estimation algorithm has been extended for application into the computational environment of the best-estimate simulation code RELAP5-3D. In control system terminology, this configuration can be thought of as a 'best-estimate' observer. The application to a distributed-parameter reactor system involves a unique modal model that approximates physical components, such as the reactor, by describing both states and parameters by an orthogonal expansion. The basic KFSLS parameter estimation is used to dynamically refine a spatially varying (distributed) parameter. The application of the distributed-parameter estimator is expected to complement a traditional nonlinear best-estimate simulation code by providing a mechanism for reducing both code input (modeling) and output (state-estimation) uncertainty in complex, distributed-parameter systems

Reactor noise analysis applications in NPP I and C systems

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

2006-07-01

Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

Development of a reactor-coolant-pump monitoring and diagnostic system. Semi-annual progress report, December 1981-May 1982

International Nuclear Information System (INIS)

Morris, D.J.; Gabler, H.C.

1982-10-01

Reactor coolant (RC) pump seal failures have resulted in excessive leakage of primary coolant into reactor containment buildings. In some cases, high levels of airborne activity and surface contamination following these failures have necessitated extensive cleanup efforts and personnel radiation exposure. Unpredictable pump seal performance has also caused forced outages and frequent maintenance. The quality of operating data has been insufficient to allow proper evaluation of theoretical RC pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables. This report describes system software and hardware development, testing, and installation work performed during the period of December 1981 through May 1982. Also described herein is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

Conceptual design of neutron diagnostic systems for fusion experimental reactor

International Nuclear Information System (INIS)

Iguchi, T.; Kaneko, J.; Nakazawa, M.

1994-01-01

Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area

International Nuclear Information System (INIS)

Yazid, M; Sutrisno; Sukarman-Aminjoyo; Zaenal-Abidin

1996-01-01

Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area has been done. The aim of this investigation is for tracing the possibility of radioactivity released in the environment during the operation of Kartini reactor. The data was evaluated were monthly monitored data taken from 1986 to 1994 period. The method of analysis was done by comparing the environmental radioactivity data before and after reactor commissioning, off side the reactor up to a radius of 5.000 meters and more than 5.000 meters from Kartini reactor and also compared to the maximum permissible radioactivity according to the current regulation. This evaluation showed that there was no indication of radioactivity release to the environment during this period of reactor operation

A basic design of SR4 instrumentation and control system for research reactor

International Nuclear Information System (INIS)

Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

2010-01-01

An SR4 instrumentation and control systems of research reactor is the equipment of nuclear research reactors as power protection devices and control systems. The equipment is to monitor safety parameters and process parameters in the state of reactor shut down, start-up, and in operation at fixed power. In the engineering of Instrumentation and control systems SR4 research reactor, its basic design consists of technical specifications of the reactor protection system devices, technical specifications of the reactor power control system devices, technical specifications information system devices, and systems process termination cabling as a support system. This basic design is used as the basis for the preparation of detailed design and subsequent engineering development of instrumentation systems and control system integrated. (author)

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)

Experience with neutron flux monitoring systems qualified for post-accident monitoring

International Nuclear Information System (INIS)

Shugars, H.G.; Miller, J.F.

1995-01-01

In this paper we discuss the environmental requirements for excore neutron flux monitors that are qualified for use during and after postulated accidents in Pressurized Water Reactors (PWRs). We emphasize PWRs designed in the United States, which are similar to those used also in parts of Western Europe and Eastern Asia. We then discuss design features of the flux monitoring systems necessary to address the environmental, functional, and regulatory requirements, and the experience with these systems. (author). 9 refs, 2 figs

On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

International Nuclear Information System (INIS)

Hawari, Ayman I.; Bourham, Mohamed A.

2010-01-01

Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

Nuclear reactor instrumentation power monitor

International Nuclear Information System (INIS)

Suzuki, Shigeru.

1989-01-01

The present invention concerns a nuclear reactor instrumentation power monitor that can be used in, for example, BWR type nuclear power plants. Signals from multi-channel detectors disposed on field units are converted respectively by LPRM signal circuits. Then, the converted signals are further converted by a multiplexer into digital signals and transmitted as serial data to a central monitor unit. The thus transmitted serial data are converted into parallel data in the signal processing section of the central monitor unit. Then, LPRM signals are taken out from each of channel detectors to conduct mathematical processing such as trip judgment or averaging. Accordingly, the field unit and the central monitor unit can be connected by way of only one data transmission cable thereby enabling to reduce the number of cables. Further, since the data are transmitted on digital form, it less undergoes effect of noises. (I.S.)

SEJV2 software package for radiation monitoring system of WWER 440 NPP

International Nuclear Information System (INIS)

Kapisovsky, V.; Jancik, O.; Kubik, I.; Bena, J.

1993-01-01

The main part of the radiation monitoring system at a WWER-440 (213 reactor type) nuclear power plant is the centralized 400-channel monitoring system 'SEJVAL' servicing twin reactor units. The SEJV2 software package is described developed to run on a PC with an IFS2 interface to the SEJVAL radiation monitoring system. It provides enhanced data presentation, record keeping and report generation, thus improving the efficiency of the health physics shift. The system was for the first time implemented at the Jaslovske Bohunice V-2 nuclear power plant with encouraging results. (Z.S.) 3 refs

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Functional safeguards for computers for protection systems for Savannah River reactors

International Nuclear Information System (INIS)

Kritz, W.R.

1977-06-01

Reactors at the Savannah River Plant have recently been equipped with a ''safety computer'' system. This system utilizes dual digital computers in a primary protection system that monitors individual fuel assembly coolant flow and temperature. The design basis for the (SRP safety) computer systems allowed for eventual failure of any input sensor or any computer component. These systems are routinely used by reactor operators with a minimum of training in computer technology. The hardware configuration and software design therefore contain safeguards so that both hardware and human failures do not cause significant loss of reactor protection. The performance of the system to date is described

On-line use of personal computers to monitor and evaluate important parameters in the research reactor DHRUVA

International Nuclear Information System (INIS)

Sharma, S.K.; Sengupta, S.N.; Darbhe, M.D.; Agarwal, S.K.

1998-01-01

The on-line use of Personal Computers in research reactors, with custom made applications for aiding the operators in analysing plant conditions under normal and abnormal situations, has become extremely popular. A system has been developed to monitor and evaluate important parameters for the research reactor DHRUVA, a 100 MW research reactor located at the Bhabha Atomic Research Centre, Trombay. The system was essentially designed for on-line computation of the following parameters: reactor thermal power, reactivity load due to Xenon, core reactivity balance and performance monitoring of shut-down devices. Apart from the on-line applications, the system has also been developed to cater some off-line applications with Local Area Network in the Dhruva complex. The microprocessor based system is designed to function as an independent unit, parallel dumping the acquired data to a PC for application programmes. The user interface on the personal computer is menu driven application software written in 'C' language. The main input parameters required for carrying out the options given in the above menu are: Reactor power, Moderator level, Coolant inlet temperature to the core, Secondary coolant flow rate, temperature rise of secondary coolant across the heat exchangers, heavy water level in the Dump tank and Drop time of individual shut off rods. (author)

Cable condition monitoring in a pressurized water reactor environment

International Nuclear Information System (INIS)

Al-Hussaini, T.J.

1988-01-01

Oconee Nuclear Station is the first nuclear plant designed, engineered and constructed by Duke Power Company. Even though the accelerated aging method was available to determine the life expectancy of the cable used in the reactor building, no natural aging data was available at that time. In order to be able to verify the condition of the reactor building cable over the life of the plant, an on-going cable monitoring plan was instituted. Various types of cable were selected to be monitored, and they were installed in cable life evaluation circuits in the reactor building. At five year intervals over the life of the plant, cable samples would be removed from these cable life evaluation circuits and tested to determine the effects of the reactor building environment on the integrity of the cable. A review of the cable life evaluation circuits and the results of the evaluation program to date is presented

Demonstration of the importance of a dedicated neutron beam monitoring system for BNCT facility

International Nuclear Information System (INIS)

Chao, Der-Sheng; Liu, Yuan-Hao; Jiang, Shiang-Huei

2016-01-01

The neutron beam monitoring system is indispensable to BNCT facility in order to achieve an accurate patient dose delivery. The neutron beam monitoring of a reactor-based BNCT (RB-BNCT) facility can be implemented through the instrumentation and control system of a reactor provided that the reactor power level remains constant during reactor operation. However, since the neutron flux in reactor core is highly correlative to complicated reactor kinetics resulting from such as fuel depletion, poison production, and control blade movement, some extent of variation may occur in the spatial distribution of neutron flux in reactor core. Therefore, a dedicated neutron beam monitoring system is needed to be installed in the vicinity of the beam path close to the beam exit of the RB-BNCT facility, where it can measure the BNCT beam intensity as closely as possible and be free from the influence of the objects present around the beam exit. In this study, in order to demonstrate the importance of a dedicated BNCT neutron beam monitoring system, the signals originating from the two in-core neutron detectors installed at THOR were extracted and compared with the three dedicated neutron beam monitors of the THOR BNCT facility. The correlation of the readings between the in-core neutron detectors and the BNCT neutron beam monitors was established to evaluate the improvable quality of the beam intensity measurement inferred by the in-core neutron detectors. In 29 sampled intervals within 16 days of measurement, the fluctuations in the mean value of the normalized ratios between readings of the three BNCT neutron beam monitors lay within 0.2%. However, the normalized ratios of readings of the two in-core neutron detectors to one of the BNCT neutron beam monitors show great fluctuations of 5.9% and 17.5%, respectively. - Highlights: • Two in-core neutron detectors and three BNCT neutron beam monitors were compared. • BNCT neutron beam monitors improve the stability in neutron

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

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

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

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

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

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

Mimic of OSU research reactor

International Nuclear Information System (INIS)

Lu, Hong; Miller, D.W.

1991-01-01

The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

EDGAR, a new plant radiation monitoring system

International Nuclear Information System (INIS)

Vuong, Q.M.; Da Costa Vieira, D.

2004-01-01

The EDGAR system is a new radiation monitoring system for nuclear power plant, reprocessing plant and nuclear research reactor for radioactive contamination, gamma and neutron field monitoring. Developed by French Atomic Energy Agency, this system provides not only complete functions of standard RMS, also allows spectroscopy level detection of alpha and beta particles based on a patented collimator unit. A complete computerized approach has been taken allowing full installation control in a single PC based display and communication unit. (author)

Development of telerobotic systems for reactor decommissioning, (2)

International Nuclear Information System (INIS)

Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

1991-01-01

This paper describes the prototype heavy-duty telerobotic system constructed as a cold test facility for the development of robotic remote handling system technology in reactor decommissioning. The total system is built up with a multi-functional electrical manipulator system, a manipulator transporter system equipped with a tripedal support mechanism, a monitoring system comprising a 3-D TV monitor, and a computer control system for overall system operation. The manipulator system consists of two master manipulators and two corresponding amphibious slave manipulators with load capacities of 100 and 25 daN, respectively. Valuable engineering experiences for developing more advanced heavy-duty telerobotic system have been gained through designing, constructing and testing the system. (author)

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Monitoring device for the stability of a reactor core

International Nuclear Information System (INIS)

Sakurai, Mikio; Yamauchi, Koki.

1983-01-01

Purpose: To avoid unnecessary limitation on the operation conditions for maintaining the reactor stability. Constitution: The reactor stability is judged by taking notice of the axial power distribution of the reactor and monitoring the same online. Specifically, signals are received from a plurality of local power distribution detectors arranged axially in the reactor core to calculate the axial power distribution in computer. Further, a certain distance L is set from the lower end of the reactor core and the total value S1 for the power distribution in the region below the set value L and the total value S2 for the region above the set value L are determined based on the thus calculated power distribution, to thereby determine the ratio: R = S1/S2 between them. Separately, a certain value r is previously determined based on analysis or experiment such as the result of operation. Then, R and r are compared in a comparator and an alarm is generated, if R >r, with respect to the stability. Since monitoring is made based on the actual index, the applicable range of the operation region can be extended. (Ikeda, J.)

Modern diagnostic systems for loose parts, vibration and leakage monitoring

International Nuclear Information System (INIS)

Kunze, U.

1997-01-01

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

Evaluation of the RSG-GAS Alpha-Beta Aerosol Contaminant Monitor Performance Under Reactor Operation Condition

International Nuclear Information System (INIS)

Hartoyo, Unggul; Setiawanto, Anto; Sumarno, Yulius

2000-01-01

Analysis to evaluate the RSG-GAS alpha-beta aerosol contaminant monitor performance was done. The high potential radiation working area such as in RSG-GAS is important to monitored for personal safety. Further it is necessary to assure that the system monitor is reliable enough under normal conditions as well as emergency condition. The method uses in this analysis are monitoring and comparing with the standard source. The standard course indicator and panel in main control room indicate that the result is 1 x 110 exp 9 Ci/m exp 3. Based on data monitor observation, the RSG-GAS alpha-beta aerosol contaminant monitor system under reactor operation condition has a good enough performance

Longtime radionuclide monitoring in the vicinity of Salaspils nuclear reactor; Dauerhaftes Monitoring der Radionuklide in Umgebung von Salaspils Kernreaktor

Energy Technology Data Exchange (ETDEWEB)

Riekstina, D.; Berzins, J.; Krasta, T. [Latvia Univ. (Latvia). Inst. of Solid State Physics; Skrypnik, O.; Alksnis, J. [Latvia Univ. (Latvia). Inst. of Chemical Physics

2016-07-01

The research nuclear reactor in Salaspils was decommissioned in 1998. Now reactor is partially dismantled and its territory is used as a temporary storage of radioactivity contaminated materials and water. Environment radioactivity monitoring for presence of artificial radionuclides in the vicinity of Salaspils nuclear reactor is carried out since 1990. Data include Cs-137 concentration in soils, tritium concentration in ground water, as well as H-3, Cs-137, Co-60 concentration and gross beta-activity of reactors sewage and rainwater drainage. The systematic monitoring allowed to detect in December 2014 a leakage from the special wastewater basin and so to prevent a pollution of ground water outside reactors territory.

Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

International Nuclear Information System (INIS)

Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

2010-01-01

The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

Performance Test for Neutron Detector and Associated System using Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

Energy Technology Data Exchange (ETDEWEB)

Bernstein, A; Bowden, N; Misner, A; Palmer, T

2007-06-27

In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7 days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

Use of digital computers in the protection system for Savannah River reactors

International Nuclear Information System (INIS)

Gimmy, K.L.

1977-06-01

Each production reactor at the Savannah River Plant has recently been provided with a protective system using dual digital computers. The dual ''safety computers'' monitor coolant temperature and flow in each of the 600 fuel assemblies in the reactor. The system provides alarms and automatic reactor shutdown (SCRAM) if these variables exceed predetermined setpoints. The system provides the primary protection for unwanted local or general power increase or assembly coolant flow reduction. Standard process control computers are used and all scanning, data output, and protective action are controlled by software prepared by Du Pont

The design and installation of a core discharge monitor for CANDU-type reactors

International Nuclear Information System (INIS)

Halbig, J.K.; Monticone, A.C.; Ksiezak, L.; Smiltnieks, V.

1990-01-01

A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses (γ,n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs

Acoustic monitoring systems tests at Indian Point Unit 1. Final report

International Nuclear Information System (INIS)

Smith, J.R.; Rao, G.V.; Craig, J.

1979-12-01

This report describes the results of a program to test acoustic monitoring systems on Indian Point Unit No. 1 under actual plant operating conditions, less the reactor core. The two types of systems evaluated were the monitoring of acoustic emissions generated by growing flaws and the monitoring of acoustic signals from leaks

Data acquisition system for nuclear reactor environment

International Nuclear Information System (INIS)

Tiwari, Akash; Tiwari, Railesha; Tiwari, S.S.; Panday, Lokesh; Suri, Nitin; Chouksey, Abhsek; Singh, Sarvesh Kumar; Dwivedi, Tarun; Agrawal, Ashish; Pandey, Pranav Kumar; Sharma, Brijnandan; Bhatia, Chirag

2004-01-01

We have designed an online real time data acquisition system for nuclear reactor environment monitoring. Data acquisition system has eight channels of analog signals and one channel of pulsed input signal from detectors like GM Tube, or any other similar input. Connectivity between the data acquisition system and environmental parameters monitoring computer is made through a wireless data communication link of 151 MHz/100 mW RF power and 10 km maximum communication range for remote data telemetry. Sensors used are gamma ionizing radiation sensor made from CsI:Tl scintillator, atmospheric pressure sensor with +/-0.1 mbar precision, temperature sensor with +/-l milli degree Celsius precision, relative humidity with +/-0.1RH precision, pulse counts with +/-1 count in 0-10000 Hz count rate measurement precision and +/-1 count is accumulated count measurement precision. The entire data acquisition system and wireless telemetry system is 9 V battery powered and the device is to be fitted on a wireless controlled mobile robot for scanning the nuclear reactor zone from remote. Wireless video camera has been planned for integration into the existing system on a later date for moving the robotics environmental data acquisition system beyond human vision reach. System development cost is Rs.25 Lacs and has been developed for Department of Atomic Energy, Government of India and Indian Defense use. (author)

Plan for Demonstration of Online Monitoring for the Light Water Reactor Sustainability Online Monitoring Project

Energy Technology Data Exchange (ETDEWEB)

Magdy S. Tawfik; Vivek Agarwal; Nancy J. Lybeck

2011-09-01

Condition based online monitoring technologies and development of diagnostic and prognostic methodologies have drawn tremendous interest in the nuclear industry. It has become important to identify and resolve problems with structures, systems, and components (SSCs) to ensure plant safety, efficiency, and immunity to accidents in the aging fleet of reactors. The Machine Condition Monitoring (MCM) test bed at INL will be used to demonstrate the effectiveness to advancement in online monitoring, sensors, diagnostic and prognostic technologies on a pilot-scale plant that mimics the hydraulics of a nuclear plant. As part of this research project, INL will research available prognostics architectures and their suitability for deployment in a nuclear power plant. In addition, INL will provide recommendation to improve the existing diagnostic and prognostic architectures based on the experimental analysis performed on the MCM test bed.

Architectural conceptual definition of the CAREM-25 reactor's control system

International Nuclear Information System (INIS)

Perez, J.C.; Santome, D.; Drexler, J.; Escudero, S.

1990-01-01

This work presents the conceptual definition of the CAREM 25 reactor's digital and monitoring control system structure. The requirements of the system are analyzed and different implementation alternatives are studied where possible basic architectures of the system and its topology are considered and evaluated. (Author) [es

Study and application of digital physical start-up system for nuclear reactor

International Nuclear Information System (INIS)

Qu Ronghong; Li Baoxiang; Xu Xiaolin

2004-01-01

The digital physical start-up system for nuclear reactor is introduced. The system was used successfully in physical start-up experiment of 10 MW high-temperature gas-cooled reactor. It is proved practically that the system not only runs reliably and calculates both rapidly and correctly and relieves the loads of operators, but also has the better characters of monitoring and showing the real-time results of experiments than the analog systems. (author)

New tritium monitor for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Jalbert, R.A.

1985-01-01

At DT-fueled fusion reactors, there will be a need for tritium monitors that can simultaneously measure in real time the concentrations of HTO, HT and the activated air produced by fusion neutrons. Such a monitor has been developed, tested and delivered to the Princeton Plasma Physics Laboratory for use at the Tokamak Fusion Test Reactor (TFTR). It uses semipermeable membranes to achieve the removal of HTO from the sampled air for monitoring and a catalyst to convert the HT to HTO, also for removal and monitoring. The remaining air, devoid of tritium, is routed to a third detector for monitoring the activated air. The sensitivities are those that would be expected from tritium instruments employing conventional flow-through ionization chambers: 1 to 3 μCi/m 3 . Its discriminating ability is approximately 10 -3 for any of the three components (HTO, HT and activated air) in any of the other two channels. For instance, the concentration of HT in the HTO channel is 10 -3 times its original concentration in the sampled air. This will meet the needs of TFTR

An integrated monitoring and control system for THOR

International Nuclear Information System (INIS)

Chou, H.P.; Chou, T.H.; Chen, T.L.

1992-01-01

The paper presents a computerized monitoring and control system for the THOR. The system is used to assist reactor operation and to facilitate data acquisition and teaching for reactor experimental laboratory courses. The design applies digital data processing for neutron detector and area monitor measurements. Signal validation is used to improve signal reliability. Color pictures in the forms of analog meters, strip charts, and bar graphs are displayed for the control room and for off-site as well. Power control is based on the 'reactivity constraint' approach for wide range adjustment and on-off logics for narrow range regulation. Algorithms are coded in C language and implemented into a 32-bit microcomputer. Evaluations have shown satisfactory results for operation and teaching needs. (author)

Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

Energy Technology Data Exchange (ETDEWEB)

Tavera D, L; Camacho L, M E

1991-01-15

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

Nuclear power plant monitoring method by neural network and its application to actual nuclear reactor

International Nuclear Information System (INIS)

Nabeshima, Kunihiko; Suzuki, Katsuo; Shinohara, Yoshikuni; Tuerkcan, E.

1995-11-01

In this paper, the anomaly detection method for nuclear power plant monitoring and its program are described by using a neural network approach, which is based on the deviation between measured signals and output signals of neural network model. The neural network used in this study has three layered auto-associative network with 12 input/output, and backpropagation algorithm is adopted for learning. Furthermore, to obtain better dynamical model of the reactor plant, a new learning technique was developed in which the learning process of the present neural network is divided into initial and adaptive learning modes. The test results at the actual nuclear reactor shows that the neural network plant monitoring system is successfull in detecting in real-time the symptom of small anomaly over a wide power range including reactor start-up, shut-down and stationary operation. (author)

Online monitoring of steam/water chemistry of a fast breeder test reactor

International Nuclear Information System (INIS)

Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

2005-01-01

Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

International Nuclear Information System (INIS)

Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

2017-01-01

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

Energy Technology Data Exchange (ETDEWEB)

Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

2017-11-01

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

Monitoring of primary circuit and reactor of NPP A-1

International Nuclear Information System (INIS)

Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

2005-01-01

Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Monitoring of homogeneity of fuel compacts for high-temperature reactors

International Nuclear Information System (INIS)

Mottet, P.; Guery, M.; Chegne, J.

Apparatus using either gamma transmission or gamma scintillation spectrometry (with NaI(Tl) detector) was developed for monitoring the homogeneity of distribution of fissile and fertile particles in fuel compacts for high-temperature reactors. Three methods were studied: Longitudinal gamma transmission which gives a total distribution curve of heavy metals (U and Th); gamma spectrometry with a well type scintillator, which rapidly gives the U and Th count rates per fraction of compact; and longitudinal gamma spectrometry, giving axial distribution curves for uranium and thorium; apparatus with four scintillators and optimization of the parameters for the measurement, permitting significantly decreasing the duration of the monitoring. These relatively simple procedures should facilitate the industrial monitoring of high-temperature reactor fuel

Reliability modeling of Clinch River breeder reactor electrical shutdown systems

International Nuclear Information System (INIS)

Schatz, R.A.; Duetsch, K.L.

1974-01-01

The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

Application Software for the Cabinet Operator Module of the Reactor Protection System

International Nuclear Information System (INIS)

Lee, Hyun-Chul; Jung, Hae-Won; Lee, Sung-Jin; Koo, Young-Ho; Kim, Seong-Tae; Kwak, Tae-Kil; Jin, Kyo-Hong

2006-01-01

A reactor protection system (RPS) plays the roles of generating the reactor trip signal and the engineered safety features (ESF) actuation signal when the monitored plant processes reach predefined limits. A Korean project group, so-called KNICS (Korean Nuclear I and C System), is developing a new digitalized RPS and the Cabinet Operator Module (COM) of the RPS which is used for the RPS integrity testing and monitoring by equipment operators. A flat panel display (FPD) with a touch screen capability is provided as a main user interface for the RPS. This paper shows the application software developed for the COM FPD. Equipment operators can monitor the status of the RPS and carry out various tests to verify system functions by means of the application software. A qualified hardware and software development environment are used to develop the application software

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

Calibration of a He accumulation fluence monitor for fast reactor dosimetry

Energy Technology Data Exchange (ETDEWEB)

Ito, Chikara [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-03-01

The helium accumulation fluence monitor (HAFM) has been developed for a fast reactor dosimetry. The HAFM measurement system was calibrated using He gas and He implanted samples and the measurement accuracy was confirmed to be less than 5%. Based on the preliminary irradiation test in JOYO, the measured He in the {sup 10}B type HAFM agreed well with the calculated values using the JENDL-3.2 library. (author)

On-line core monitoring system based on buckling corrected modified one group model

International Nuclear Information System (INIS)

Freire, Fernando S.

2011-01-01

Nuclear power reactors require core monitoring during plant operation. To provide safe, clean and reliable core continuously evaluate core conditions. Currently, the reactor core monitoring process is carried out by nuclear code systems that together with data from plant instrumentation, such as, thermocouples, ex-core detectors and fixed or moveable In-core detectors, can easily predict and monitor a variety of plant conditions. Typically, the standard nodal methods can be found on the heart of such nuclear monitoring code systems. However, standard nodal methods require large computer running times when compared with standards course-mesh finite difference schemes. Unfortunately, classic finite-difference models require a fine mesh reactor core representation. To override this unlikely model characteristic we can usually use the classic modified one group model to take some account for the main core neutronic behavior. In this model a course-mesh core representation can be easily evaluated with a crude treatment of thermal neutrons leakage. In this work, an improvement made on classic modified one group model based on a buckling thermal correction was used to obtain a fast, accurate and reliable core monitoring system methodology for future applications, providing a powerful tool for core monitoring process. (author)

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dawn Scates

2010-10-01

A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

Power distribution monitor for nuclear reactor

International Nuclear Information System (INIS)

Nishizawa, Yasuo; Kiguchi, Takashi.

1974-01-01

Object: To compare the measured local power region monitor (LPRM) index with the result of a primary calculation to correct the threshold condition for the primary calculation thereby to rapidly grasp and monitor the existing power distribution. Structure: The index of an LPRM disposed in a nuclear reactor is processed in a data processor to remove therefrom a noise, and transmitted to a threshold condition processor to be stored therein. The LPRM index measured by the threshold condition processor is compared with the calculated LPRM value transmitted from the primary processor, whereby the threshold condition is corrected and transmitted to the primary processor. After the completion of calculation, the traversing incore probe (TIP) indexing value is converted to a thermal output distribution or a linear output density distribution and transmitted to an output indicator or an output typewriter. The operator may monitor the existing power distribution by monitoring the output indicator. (Kamimura, M.)

State-of-the-art incore detector system provides operational and safety benefits: Example, Hanford N Reactor

International Nuclear Information System (INIS)

Toffer, H.

1988-08-01

A presentation on the operational and safety benefits that can be derived from a state-of-the-art incore neutron monitoring system has been prepared for the DOE/ANL training course on ''The Potential Safety Impact of New and Emerging Technologies on the Operation of DOE Nuclear Facilities.'' Advanced incore neutron flux monitoring systems have been installed in some commercial reactors and should be considered for any new reactor designs or as backfits to existing plants. The recent installation of such a system at the Hanford N Reactor is used as an example in this presentation. Unfortunately, N Reactor has been placed in a cold standby condition and the full core incore system has not been tested under power conditions. Nevertheless, the evaluations that preceded the installation of the full core system provide interesting insight into the operational and safety benefits that could be expected

Reactor Emergency Action Level Monitor: Knowledge acquisition experiences

International Nuclear Information System (INIS)

Touchton, R.A.

1987-01-01

This paper presents the Knowledge Acquisition experiences in developing the Reactor Emergency Action Level Monitor (REALM) Expert System Prototype. REALM is an expert system which interprets plant sensor data and provides advice on the proper emergency classification. The REALM project is being funded by the Electric Power Research Institute, Consolidated Edison is serving as the host utility, and the effort is being conducted by Technology Applications, Inc. REALM is being designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity, ultimately operating in a real-time, on-line processing environment. The paper discusses briefly the direct knowledge acquisition techniques used by the project team (who are themselves power industry engineers), to extract relevant knowledge from plant specifications and procedures

Modernization incore monitoring system of WWER-1000 reactors (V-320)

International Nuclear Information System (INIS)

Mitin, Valentin; Semchenkov, Yurij; Kalinushkin, Andrey

2008-01-01

Modern ICIS system for VVER-1000, including a number of sensors, cable runs, corresponding measuring equipment and computer engineering, software, accumulated 30 year experience of interaction researches on VVER reactors and is capable to ensure carrying out of control, protection, informational, diagnostic functions and thus to promote real increase of quality, reliability and safety in nuclear fuel and NPP power units operation

The nuclear reactor remote monitoring system as an instrument for early detection and evaluation of incidents or accidents

International Nuclear Information System (INIS)

Eder, E.

1981-01-01

The remote monitoring system for nuclear reactors (KFUe) in Bavaria is a technical means to support the supervising authority. The measuring devices installed in the KFUe make it possible to perceive very early such accidents which could be liable for unplanned releases of greater amounts of radioactive substances. The data compiled by the KFUe will be also used as basis for further statutory supervisions, if necessary. When alarm thresholds in the measuring devices are exceeded the stand-by supervision staff on the Bavarian State Office for Environmental Protection will be informed automatically. Further decisions will be met by the supervision staff. (orig.) [de

Method and apparatus for monitoring the axial power distribution within the core of a nuclear reactor, exterior of the reactor

International Nuclear Information System (INIS)

Graham, K.F.; Gopal, R.

1978-01-01

A method and apparatus for establishing the axial flux distribution of a reactor core from monitored responses obtained exterior of the reactor is described. The monitored responses are obtained from at least three axially spaced flux responsive detectors that are positioned within proximity of the periphery of the reactor core. The detectors provide corresponding electrical outputs representative of the flux monitored. The axial height of the core is figuratively divided at a plurality of space coordinates sufficient to provide reconstruction in point representation of the relative flux shape along the core axis. The relative value of flux at each of the spaced coordinates is then established from a sum of the electrical outputs of the detectors, respectively, algebraically modified by a corresponding preestablished constant

Remote Monitoring and Instrumentation Strategies for Integral Reactors

International Nuclear Information System (INIS)

Upadhyaya, Belle R.; Lish, Matthew R.; Tarver, Rayan A.; Hines, J. Wesley

2014-01-01

The University of Tennessee is engaged in research and development projects related to instrumentation and controls for small modular reactors (SMR) and integral pressurized water reactors (iPWR). The approach incorporates the deployment of physics-based models for control design and parameter estimation, development of noncontact sensors for flow monitoring, and placement of sensors to maximize fault detection and isolation. The results of research and development illustrate the feasibility of sensor location in space-constrained environment. Major issues and challenges in I and C design are addressed

Remote Monitoring and Instrumentation Strategies for Integral Reactors

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, Belle R.; Lish, Matthew R.; Tarver, Rayan A.; Hines, J. Wesley [University of Tennessee, Knoxville (United States)

2014-08-15

The University of Tennessee is engaged in research and development projects related to instrumentation and controls for small modular reactors (SMR) and integral pressurized water reactors (iPWR). The approach incorporates the deployment of physics-based models for control design and parameter estimation, development of noncontact sensors for flow monitoring, and placement of sensors to maximize fault detection and isolation. The results of research and development illustrate the feasibility of sensor location in space-constrained environment. Major issues and challenges in I and C design are addressed.

Humos monitoring system of leaks in to the containment atmosphere

International Nuclear Information System (INIS)

Matal, O.; Zaloudek, J.; Matal, O. Jr.; Klinga, J.; Brom, J.

1997-01-01

HUmidity MOnitoring System (HUMOS) has been developed and designed to detect the presence of leak in selected primary circuit high energy pipelines and components that are evaluated from the point of view of Leak Before Break (LBB) requirements. It also requires to apply technical tools for detection and identification of coolant leaks from primary circuit and components of PWRs reactors. Safety significant of leaks depend on: leak source (location); leak rate, and leak duration. Therefore to detect and monitor coolant leaks in to the containment atmosphere during reactor operation is one of important safety measures. As potential leak sources flange connection in the upper head region of WWER reactors can be considered. HUMOS does not rely on the release of radioactivity to detect leaks but rather the relies on detection of moisture in the air resulting from a primary boundary leak. Because HUMOS relies on moisture and temperature detection, leaks can be detected without requiring the reactor to be critical. Therefore leaks can be detected during integrity pressure tests and any other mode of operation provided the reactor ventilation system is operating and primary circuit and components are pressurized. 3 figs

Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Pilgrim Nuclear Power Station

International Nuclear Information System (INIS)

Selan, J.C.

1982-01-01

This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Pilgrim Nuclear Power Station. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

International Nuclear Information System (INIS)

Kawamura, Makoto; Takai, Katsuaki.

1985-01-01

Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

Development of a reactor coolant pump monitoring and diagnostic system. Progress report, June 1982-July 1983

International Nuclear Information System (INIS)

Morris, D.J.; Sommerfield, G.A.

1983-12-01

The quality of operating data has been insufficient to allow proper evaluation of theoretical reactor coolant (RC) pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables: The rotordynamic behavior of the pump shaft and related components, the internal conditions and performance of the seals, and the plant or pump operating environment (controlled by the plant operator). Interrelationships between these areas will be developed during the data collection task, scheduled to begin in October 1983 (for a full fuel cycle at Davis-Besse). This report describes system software and hardware development, testing, and installation work performed during this period. Also described is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

NSSS Component Control System Design of Integral Reactor

International Nuclear Information System (INIS)

Lee, Joon Koo; Kwon, Ho Je; Jeong, Kwong Il; Park, Heui Youn; Koo, In Soo

2005-01-01

MMIS(Man Machine Interface System) of an integral reactor is composed of a Control Room, Plant Protection System, Control System and Monitoring System which are related with the overall plant operation. MMIS is being developed with a new design concept and digital technology to reduce the Human Factor Error and improve the systems' safety, reliability and availability. And CCS(component control system) is also being developed with a new design concept and digital hardware technology A fully digitalized system and design concept are introduced in the NSSS CCS

Concept for creating program-technical complex of safety monitoring with system of safety parameters presentation functions on the basis of routine WWER-1000 systems

International Nuclear Information System (INIS)

Dunaev, V.G.; Tarasov, M. V.; Povarov, P.V.

2005-01-01

Prerequisites of creating the software-hardware complex for reactor safety monitoring on the Volgodonsk NPP are analyzed and generalized. The concept of this complex is based on functions of the safety parameters presentation system. It will serve as an interface between operator and technological process and give to operator a possibility to estimate quickly the state of the safety of the nuclear power unit. The complex will be created on the basis of routine reactor monitoring and control systems intended for the WWER-1000 reactor. In addition to existing soft- and hard-wares for reactor monitoring and for analysis of technological archive, it is proposed to create and connect in parallel the new software-hardware complex which ensures calculation and presentation of generalized factors of reactor safety [ru

Fort St. Vrain high temperature gas-cooled reactor. Pt. 12. The dew point moisture monitor testing program

Energy Technology Data Exchange (ETDEWEB)

Olson, H.G. (Colorado State Univ., Fort Collins (USA). Dept. of Mechanical Engineering); Brey, H.L. (Public Service Co. of Colorado, Denver (USA)); Swart, F.E. (Gas-Cooled Reactor Associates, La Jolla, CA (USA)); Forbis, J.M. (Storage Technology Corp., Louisville, CO (USA))

1982-09-01

Moisture ingress into the core volume could cause damaging reactions with the moderator-reflector graphite and burnable poison, therefore a dew point moisture monitoring system has been developed with the basic design criteria that a plant protective system trip is signaled after the system detects high primary coolant helium moisture levels and that the system is able to correctly identify which of two steam generator loops is leaking. Modifications to the sample supplies to the monitors were necessary to reduce the system's unsatisfactory response time at lower reactor power levels.

Implementation of digital control and protection systems of China advanced research reactor

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

Experience with digital acoustic monitoring systems for PWRs and BWRs

International Nuclear Information System (INIS)

Olma, B.J.

1998-01-01

Substantial progress could be reached both in system technics and in application of digital acoustic monitoring systems for assessing mechanical integrity of reactor primary systems. For the surveillance of PWRs and BWRs during power operation of the plants, acoustic signals of Loose Parts Monitoring System sensors are continuously monitored for signal bursts associated with metallic impacts. ISTec/GRS experience with its digital systems MEDEA and RAMSES has shown that acoustic signature analysis is very successful for detecting component failures at an early stage. Methods for trending and classification of digital burst signals are shown, experience with their practical use will be presented. (author)

1996 environmental monitoring report for the Naval Reactors Facility

International Nuclear Information System (INIS)

1996-01-01

The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

Experience and evaluation of advanced on-line core monitoring system 'BEACON' at IKATA site

International Nuclear Information System (INIS)

Fujitsuka, Nobumichi; Tanouchi, Hideyuki; Imamura, Yasuhiro; Mizobuchil, Daisuke

1997-01-01

Shikoku Electric Power Company installed BEACON core monitoring system into IKATA unit 3 in May 1994. During its first cycle of core operation, various operational data were obtained including data of some anomalous reactor conditions introduced for the test objective of the plant start-up. This paper presents the evaluation of the BEACON system capability based on this experience. The system functions such as core monitoring and anomaly detection, prediction of future reactor conditions and increased efficiency of core management activities are discussed. Our future plan to utilize the system is also presented. (authors)

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)

Identification of process dynamics. Stability monitoring in BWR type reactors

International Nuclear Information System (INIS)

Abrahamsson, P.; Hallgren, P.

1991-06-01

Identification of process dynamics is used for stability monitoring in nuclear reactors (Boiling Water Reactor). This report treats the problem of estimating a damping factor and a resonance frequency from the neutron flux as measured in the reactor. A new parametric online method for identification is derived and presented, and is shown to meet the requirements of stability monitoring. The technique for estimating the process parameters is based on a recursive lattice filter algorithm. The problem of time varying parameters and offset, as well as offline experiments and signal processing are treated. All parts are implemented in a realtime program, using the language C. In comparison with earlier identifications, the new way of estimating the damping factor is shown to work well. Estimates of both the damping factor and the resonance frequency show a stable and reliable behavior. Future development and improvements are also indicated. (au)

The development of a new steam generator leak monitoring system

International Nuclear Information System (INIS)

Ding Shengyao; Xu Kun; Huang Xiaojian; Wang Peiliang; Nie Jing

2006-01-01

The principle idea for the monitoring system is based on the time is different when radiation nuclei 16 N and 19 O travel the different distances from reactor center via hot, bend and cold point of U-tube in steam generator. Because of the decay time T 1/2 for 16 N and 19 O are different, the ratios of 16 N and 19 O activities are different too. By using the different ratio, the authors can obtain the leak location of U-tube. The radioactivities and their ratio of 16 N to 19 O in our swimming pool reactor were measurement by use the monitoring system, the measured results show its quality is reliable. (authors)

Diagnostic system for primary circuits of pressurized-water reactors

International Nuclear Information System (INIS)

Liska, J.; Majer, J.

1983-01-01

The diagnostic system monitors the reactor, the main circulating pipe, the main circulating pump, the main shut-off valve, the steam generator and the pressurizer. Diagnostic signals are obtained from the sensors designed for operation measurements and from sensors for special diagnostic purposes. The following operations are carried out: detection of dangerous dynamic stress of components, detection of damage to functional surfaces of components, detection of occurrence and propagation of defects in component materials, detection of loose particles and foreign bodies, detection of coolant leakage, detection of coolant boiling in the core and detection of impermissible non-homogeneities of fields of physical quantities in the core. The diagnostic system comprises: monitoring, classification of properly investigated effects, periodical tracing and long-term tracing. The operational diagnostics system developed by the SKODA Concern consists of a vibration monitoring system, a spectral analysis system and a central evaluation system. (M.D.)

Application study of EPICS-based redundant method for reactor control system

International Nuclear Information System (INIS)

Zhang Ning; Han Lifeng; Chen Yongzhong; Guo Bing; Yin Congcong

2013-01-01

In the reactor control system prototype development of TMSR (Thorium Molten Salt Reactor system, CAS) project, EPICS (Experimental Physics and Industrial Control System) is adopted as Instrument and Control software platform. For the aim of IOC (Input/Output Controller) redundancy and data synchronization of the system, the EPICS-based RMT (Redundancy Monitor Task ) software package and its data-synchronization component CCE (Continuous Control Executive) were introduced. By the development of related IOC driver, redundant switch-over control of server IOC was implemented. The method of redundancy implementation using RMT in server and redundancy performance test for power control system are discussed in this paper. (authors)

Renewal of reactor cooling system of JMTR. Reactor building site

International Nuclear Information System (INIS)

Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

2012-03-01

The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

Distributed computer control system for reactor optimization

International Nuclear Information System (INIS)

Williams, A.H.

1983-01-01

At the Oldbury power station a prototype distributed computer control system has been installed. This system is designed to support research and development into improved reactor temperature control methods. This work will lead to the development and demonstration of new optimal control systems for improvement of plant efficiency and increase of generated output. The system can collect plant data from special test instrumentation connected to dedicated scanners and from the station's existing data processing system. The system can also, via distributed microprocessor-based interface units, make adjustments to the desired reactor channel gas exit temperatures. The existing control equipment will then adjust the height of control rods to maintain operation at these temperatures. The design of the distributed system is based on extensive experience with distributed systems for direct digital control, operator display and plant monitoring. The paper describes various aspects of this system, with particular emphasis on: (1) the hierarchal system structure; (2) the modular construction of the system to facilitate installation, commissioning and testing, and to reduce maintenance to module replacement; (3) the integration of the system into the station's existing data processing system; (4) distributed microprocessor-based interfaces to the reactor controls, with extensive security facilities implemented by hardware and software; (5) data transfer using point-to-point and bussed data links; (6) man-machine communication based on VDUs with computer input push-buttons and touch-sensitive screens; and (7) the use of a software system supporting a high-level engineer-orientated programming language, at all levels in the system, together with comprehensive data link management

Method of fueling for a nuclear reactor

International Nuclear Information System (INIS)

Igarashi, Takao.

1983-01-01

Purpose: To enable the monitoring of reactor power with sufficient accuracy, upon starting even without existence of neutron source in case of a low average burnup degree in the reactor core. Constitution: Each of fuel assemblies is charged such that neutron source region monitors for the start-up system in a reactor core neutron instrumentation system having nuclear fuel assemblies and a neutron instrumentation system are surrounded with 4 or 16 fuel assemblies of a low burnup degree. Then, the average burnup degree of the fuel assemblies surrounding the neutron source region monitors are increased than the reactor core burnup degree, whereby neutrons released from the peripheral fuels are increased, sufficient number of neutron counts can be obtained even with no neutron sources upon start-up and the reactor power can be monitored at a sufficient accuracy. (Sekiya, K.)

Instrument accuracy in reactor vessel inventory tracking systems

International Nuclear Information System (INIS)

Anderson, J.L.; Anderson, R.L.; Morelock, T.C.; Hauang, T.L.; Phillips, L.E.

1986-01-01

Instrumentation needs for detection of inadequate core cooling. Studies of the Three Mile Island accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC as well as to provide more complete information for operator control of safety injection flow to minimize the consequences of such an accident. In 1980, the US Nuclear Regulatory Commission (NRC) required further studies by the industry and described ICC instrumentation design requirements that included human factors and environmental considerations. On December 10, 1982, NRC issued to Babcock and Wilcox (B and W) licensees orders for Modification of License and transmitted to pressurized water reactor licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements include upgraded subcooling margin monitors (SMM), upgraded core exit thermocouples (CET), and installation of a reactor coolant inventory tracking system. NRC Regulatory Guide 1.97, which covers accident monitoring instrumentation, was revised (Rev. 3) to be consistent with the requirements of item II.F.2 of NUREG-0737

Construction of a Vibration Monitoring System for HANARO's Rotating Machinery and Analysis of Pump Vibration Signals

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Yoon, Doo Byung

2005-01-01

HANARO is an open-tank-in-pool type research reactor with a thermal power of 30MW. In order to remove the heat generated by the reactor core and the reflector vessel, primary cooling pumps and reflector cooling pumps circulate coolant. These pumps are installed at the RCI(Reactor Concrete Island) which is covered by heavy concrete hatches. For the prevention of an abnormal operation of these pumps in the RCI, it is necessary to construct a vibration monitoring system that provides an alarm signal to the reactor control room when the rotating speed or the vibration level exceeds the allowable limit. The first objective of this work is to construct a vibration monitoring system for HANARO's rotating machinery. The second objective is to verify the possibility of condition monitoring of the rotating machinery. To construct a vibration monitoring system, as a first step, the standards and references related to the vibration monitoring system were investigated. In addition, to determine the number and the location of sensors that can effectively characterize the overall vibration of a pump, the vibration of the primary cooling pumps and the reflector cooling pumps were measured. Based on these results, detailed construction plans for the vibration monitoring system for HANARO were established. Then, in accordance with the construction plans, the vibration monitoring system for HANARO's rotating machinery was manufactured and installed at HANARO. To achieve the second objective, FFT analysis and bearing fault detection of the measured vibration signals were performed. The analysis results demonstrate that the accelerometers mounted at the bearing locations of the pumps can effectively monitor the pump condition

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

Modernization in-core monitoring system, of WWER-1000 reactors (W-320) by fuel assemblies with individual characteristics using

International Nuclear Information System (INIS)

Mitin, V. I.; Semchenkov, J. M.; Kalinushkin, A. E.

2007-01-01

The present report covers object, conception, engineering solution of construction of modern system of high-powered reactor in-core control, including WWER-1000 (V-320) reactors. It is known that ICMS main task is on-line monitoring distribution of power release field and its functioning independently of design programs to avoid common reason error. It is shown in what way field of power release recovery has been carrying on; rest on the signals of in-core neutron and temperature sensors. On the base of the obtained and refined information there have being automatically generated signals of preventive and emergency protection on local parameters (linear power to the maximum intensive fuel elements, reserve to heat exchange crisis, 'picking factor'). There have represented technology on sensors and processing methods of SPND and TC signals, ICIS composition and structure, program hard ware, system and applied software. Structure, composition and the taken decisions allow combining class 1E and class B and C tasks in accordance with international norms of separation and safety classes' realization. At present ICIS-M is a system, providing implementation of control, safety, information and diagnostic functions, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and NPP units. And at the same time it reduces human factor negative influence to core work thermo technical eliability in the operational process (Authors)

Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

International Nuclear Information System (INIS)

Pearson, G.G.; Meixler, L.D.; Sirsingh, R.A.P.

1992-01-01

This paper reports on the TFTR Tritium Stack and Area Monitoring Systems which have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. Area monitors provide surveillance of the air in various rooms at TFTR. Stack monitors monitor the air at the TFTR test site that is exhausted through the HVAC systems, from the room exhaust stacks and the tritium systems process vents. The philosophies for the implementation of the Stack and Area Tritium Monitoring Systems at TFTR are to use hardwired controls wherever personnel protection is involved, and to take advantage of modern intelligent controllers to provide a distributed system to support the functions of tracking, displaying, and archiving concentration levels of tritium for all of the monitored areas and stacks

AFRRI TRIGA Reactor water quality monitoring program

International Nuclear Information System (INIS)

Moore, Mark; George, Robert; Spence, Harry; Nguyen, John

1992-01-01

AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

Best-estimate reactor core monitor using state feedback strategies to resolve uncertainties

International Nuclear Information System (INIS)

Martin, R.P.

1997-01-01

The development and demonstration of a new algorithm for quantifying uncertainty in best-estimate simulation codes has been investigated. Demonstration is given by way of a prototype reactor core monitor. The architecture of this monitor integrates a distributed parameter estimation technique and the infrastructure required to support this control theory-based algorithm into a production-grade best-estimate simulation code. The Kalman filter with the sequential least-squares parameter estimation algorithm has been extended for application into the computational environment of a best-estimate simulation code, i.e., RELAP5/DOE. In control system terminology this configuration can be thought of as a best-estimate observer

Reactor power reduction system and method

International Nuclear Information System (INIS)

Bruno, S.J.; Dunn, S.A.; Raber, M.

1978-01-01

A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

Development of an artificial neural network for nuclear power monitoring and fault detection in the IEA-R1 research reactor at IPEN

International Nuclear Information System (INIS)

Bueno, Elaine Inacio; Ting, Daniel Kao Sun; Goncalves, Iraci M.P.

2005-01-01

The purpose of this paper is to develop a system to monitor the nuclear power of a reactor using Artificial Neural Networks. The database used in this work was developed using a theoretical model of IEA-R1 Research Reactor. The IEA-R1 is a pool type reactor of 5 MW, cooled and moderated by light water, and uses graphite and beryllium as reflector. To monitor the nuclear power the following variables were chosen: T3 . temperature above the reactor core, T4 . outlet core temperature, FE01 . primary loop flow rate and the nuclear power. The inputs are T3, T4 and FE01 and the output is the nuclear power. It was used several networks using the backpropagation algorithm. The conclusion is that the multiplayer perceptrons networks (MLPs), training by the backpropagation algorithm, can be used to solve this problem. The results obtained with the MLPs networks are satisfactory and the mean square error was in the order of 10 -4 during the network training and in the order of 10 -2 during the network testing. We intend to monitor the other variables of this model using the same methodology, and after this we will use the real database from the system to compare the results obtained with the model. The monitoring of the reactor variables is part of the development of a fault detection and isolation system which is underway and which is, by its turn, part of a comprehensive ageing management program. (author)

A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

International Nuclear Information System (INIS)

Meng Qinghu; Meng Qingfeng; Feng Wuwei

2012-01-01

Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

International Nuclear Information System (INIS)

Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

1985-02-01

Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

Reactor cover gas monitoring at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

1986-09-01

The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

Development of distributed plant monitoring and diagnosis system at Monju

International Nuclear Information System (INIS)

Okusa, Kyoichi; Tamayama, Kiyoshi; Kitamura, Tomomi

2003-01-01

In a nuclear plant, it is required to detect an anomaly as early as possible and to inhibit adverse consequences. This requirement is especially important for a prototype Fast Breeder Reactor Monju. Therefore, a monitoring and diagnosis system is required to be developed for Monju plant equipments. In these days, such a monitoring and diagnosis system can be realized using Web technology with rationalized system resources due to the remarkable progress of computer network technology. Then, we developed a Web based platform for the monitoring and diagnosis system of Monju. Distributed architecture, standardization and highly flexible system structure have been taken account of in the development. This newly developed platform and prototype monitoring and diagnosis systems have been validated. Prototype monitoring and diagnosis systems on the platform acquire Monju plant data and display the data on client computers using Monju intranet with acceptable delay times. The prototype monitoring and diagnosis systems for Monju have been developed on the platform and the whole system has been validated. (author)

N13 - based reactor coolant pressure boundary leakage system

International Nuclear Information System (INIS)

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

1980-05-01

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

Development of alarm logics for critical function monitoring in SMART-P reactor

Energy Technology Data Exchange (ETDEWEB)

Seong, Seung Hwan; Hur, Seop; Seo, Jae Kwang; Lee, Tae Ho; Park, Cheon Tae; Kang, Han Ok

2003-04-01

The alarm logics for the critical functions of SMART-P reactor are developed, which are based on the those of Korean Standard Nuclear power Plant(KSNP). The SMART-P reactor is an integral typed nuclear power plant and has the some different design features compared to those of KSNP. It, however, has the similar features in critical functions because it is a kind of pressurized water reactor. The alarm logics for Critical Function Monitoring System(CFMS) in SMART-P are developed from those for CFMS in KSNP. The alarm logics of CFMS in only the primary loop are, therefore, developed, though the general CFMS covered the status of primary and secondary loop including the features of the containment. The specific setpoint of related variables related to the alarm logics will be developed after the specific designs of SMART-P are finished. In appendix, we describe the conceptual architecture and variables of display screens on CFMS according to the developed alarm logics.

The failure diagnoses of nuclear reactor systems

International Nuclear Information System (INIS)

Sheng Huanxing.

1986-01-01

The earlier period failure diagnoses can raise the safety and efficiency of nuclear reactors. This paper first describes the process abnormality monitoring of core barrel vibration in PWR, inherent noise sources in BWR, sodium boiling in LMFBR and nuclear reactor stability. And then, describes the plant failure diagnoses of primary coolant pumps, loose parts in nuclear reactors, coolant leakage and relief valve location

Cyber security for remote monitoring and control of small reactors

Energy Technology Data Exchange (ETDEWEB)

Trask, D., E-mail: dave.trask@cnl.ca [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Jung, C. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada); MacDonald, M., E-mail: marienna.macdonald@cnl.ca [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

There is growing international interest and activity in the development of small nuclear reactor technology with a number of vendors interested in building small reactors in Canada to serve remote locations. A common theme of small reactor designs proposed for remote Canadian locations is the concept of a centrally located main control centre operating several remotely located reactors via satellite communications. This theme was echoed at a recent IAEA conference where a recommendation was made to study I&C for remotely controlled small modular reactors, including satellite links and cyber security. This paper summarizes the results of an AECL-CNSC research project to analyze satellite communication technologies used for remote monitoring and control functions in order to provide cyber security regulatory considerations. The scope of this research included a basic survey of existing satellite communications technology and its use in industrial control applications, a brief history of satellite vulnerabilities and a broad review of over 50 standards, guidelines, and regulations from recognized institutions covering safety, cyber security, and industrial communication networks including wireless communications in general. This paper concludes that satellite communications should not be arbitrarily excluded by standards or regulation from use for the remote control and monitoring of small nuclear reactors. Instead, reliance should be placed on processes that are independent of any particular technology, such as reducing risks by applying control measures and demonstrating required reliability through good design practices and testing. Ultimately, it is compliance to well-developed standards that yields the evidence to conclude whether a particular application that uses satellite communications is safe and secure. (author)

Cyber security for remote monitoring and control of small reactors

International Nuclear Information System (INIS)

Trask, D.; Jung, C.; MacDonald, M.

2014-01-01

There is growing international interest and activity in the development of small nuclear reactor technology with a number of vendors interested in building small reactors in Canada to serve remote locations. A common theme of small reactor designs proposed for remote Canadian locations is the concept of a centrally located main control centre operating several remotely located reactors via satellite communications. This theme was echoed at a recent IAEA conference where a recommendation was made to study I&C for remotely controlled small modular reactors, including satellite links and cyber security. This paper summarizes the results of an AECL-CNSC research project to analyze satellite communication technologies used for remote monitoring and control functions in order to provide cyber security regulatory considerations. The scope of this research included a basic survey of existing satellite communications technology and its use in industrial control applications, a brief history of satellite vulnerabilities and a broad review of over 50 standards, guidelines, and regulations from recognized institutions covering safety, cyber security, and industrial communication networks including wireless communications in general. This paper concludes that satellite communications should not be arbitrarily excluded by standards or regulation from use for the remote control and monitoring of small nuclear reactors. Instead, reliance should be placed on processes that are independent of any particular technology, such as reducing risks by applying control measures and demonstrating required reliability through good design practices and testing. Ultimately, it is compliance to well-developed standards that yields the evidence to conclude whether a particular application that uses satellite communications is safe and secure. (author)

Monitoring device for the power distribution within a nuclear reactor core

International Nuclear Information System (INIS)

Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

1986-01-01

Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

TARMS, an on-line boiling water reactor operation management system

International Nuclear Information System (INIS)

Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

1984-01-01

The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de

A CAMAC based real-time noise analysis system for nuclear reactors

International Nuclear Information System (INIS)

Ciftcioglu, O.

1987-01-01

A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals coverted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under porgram control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance. (orig.)

Plant monitoring and signal validation at HFIR

International Nuclear Information System (INIS)

Mullens, J.A.

1991-01-01

This paper describes a monitoring system for the Oak Ridge National Laboratory's (ORNL'S) High Flux Isotope Reactor (HFIR). HFIR is an 85 MW pressurized water reactor designed to produce isotopes and intense neutron beams. The monitoring system is described with respect to plant signals and computer system; monitoring overview; data acquisition, logging and network distribution; signal validation; status displays; reactor condition monitoring; reactor operator aids. Future work will include the addition of more plant signals, more signal validation and diagnostic capabilities, improved status display, integration of the system with the RELAP plant simulation and graphical interface, improved operator aids, and an alarm filtering system. 8 refs., 7 figs. (MB)

Monitoring PWR reactor vessel liquid level with SPNDs during LOCAs

International Nuclear Information System (INIS)

Adams, J.P.

1982-01-01

Data from in-core self-powered neutron detectors taken during two nuclear loss-of-coolant accident simulations have been correlated with core moderator density changes. The detector current attenuation has been calculated during blowdown and reflood phases of the simulation. Based on these data, it is concluded that these detectors could be used to monitor reactor vessel liquid level during loss-of-coolant accidents in pressurized water reactors

Accuracy of helium accumulation fluence monitor for fast reactor dosimetry

Energy Technology Data Exchange (ETDEWEB)

Ito, Chikara; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-03-01

A helium (He) accumulation fluence monitor (HAFM) has been developed for fast reactor dosimetry. In order to evaluate the measurement accuracy of neutron fluence by the HAFM method, the HAFMs of enriched boron (B) and beryllium (Be) were irradiated in the Fast Neutron Source Reactor `YAYOI`. The number of He atoms produced in the HAFMs were measured and compared with the calculated values. As a result of this study, it was confirmed that the neutron fluence could be measured within 5 % by the HAFM method, and that met the required accuracy for fast reactor dosimetry. (author)

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)

Primary cooling system for BWR type reactor

International Nuclear Information System (INIS)

Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

1993-01-01

The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

Mobile means for the monitoring of atmospheric contamination in a reactor building

International Nuclear Information System (INIS)

Marques, S.; Lestang, M.

2009-01-01

After having evoked the context and challenges of contamination monitoring when exploiting nuclear reactors, the authors discuss the representativeness of the atmospheric contamination measurement as it depends on the different physicochemical forms of radionuclides present in the circuits. They indicate the different gaseous or aerosol radioactive elements which are monitored within EDF installations. They discuss the incorporation of monitoring means at the installation design level, briefly present the use of beacons inside and outside the reactor building. They describe how monitoring is organized on the basis of alert threshold adjustments: an investigation threshold and an evacuation threshold. They discuss the beacon (or sensor) selection and indicate recommendations for their implementation for optimization purposes. They indicate where these beacons are installed and evoke the experimentation of networked mobile beacons with data remote transmission

Methodology of nuclear reactor monitoring and diagnostics using information dimension

International Nuclear Information System (INIS)

Suzudo, Tomoaki; Hayashi, Koji; Shinohara, Yoshikuni

1993-01-01

Reactor noise analysis method based on information dimension is applied to the monitoring and diagnosing of power oscillation. The method focuses on the utilization of the slope of the correlation integral (SOCI) which determines the information dimension of attractors. For practical application, the information dimension is expected to be the same as the fractal dimension of attractors; it can be used to classify different asymptotic regimes of nonlinear dynamical systems. We examined a real power oscillation using SOCI and the results implied that the oscillation was just a noisy limit cycle, although it is not possible to assert that there is no chaotic character in the oscillation because large oscillatory time-series data sets are not available. In addition, the application of SOCI to the real-time monitoring of power oscillation is proposed and examined. (author)

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

A novel method for in-situ estimation of time constant for core temperature monitoring thermocouples of operating reactors

International Nuclear Information System (INIS)

Sylvia, J.I.; Chandar, S. Clement Ravi; Velusamy, K.

2014-01-01

Highlights: • Core temperature sensor was mathematically modeled. • Ramp signal generated during reactor operating condition is used. • Procedure and methodology has been demonstrated by applying it to FBTR. • Same technique will be implemented for all fast reactors. - Abstract: Core temperature monitoring system is an important component of reactor protection system in the current generation fast reactors. In this system, multiple thermocouples are housed inside a thermowell of fuel subassemblies. Response time of the thermocouple assembly forms an important input for safety analysis of fast reactor and hence frequent calibration/time constant estimation is essential. In fast reactors the central fuel subassembly is provided with bare fast response thermocouples to detect under cooling events in reactor and take proper safety action. On the other hand, thermocouples in thermowell are mainly used for blockage detection in individual fuel subassemblies. The time constant of thermocouples in thermowell can drift due to creep, vibration and thermal fatigue of the thermowell assembly. A novel method for in-situ estimation of time constant is proposed. This method uses the Safety Control Rod Accelerated Mechanism (SCRAM) or lowering of control Rod (LOR) signals of the reactor along with response of the central subassembly thermocouples as reference data. Validation of the procedure has been demonstrated by applying it to FBTR

Seismic monitoring of the Creys-Malville plant - Problems raised by the seismic behaviour of a fast breeder reactor

International Nuclear Information System (INIS)

Descleve, P.; Barrau, P.

1988-01-01

CREYS-MALVILLE reached full power in December 1986 and is presently the largest sodium cooled reactor in operation. Well established procedures of safety evaluation have been used for the design but for a large size reactor special attention must be paid to the effects of seismic disturbances. This paper describes the seismic protection and monitoring system of the plant, the core behaviour which is specific to fast reactors and the test performed to verify the analyses. Finally the seismic impact on the construction can be established as an indication for future plants. (author)

Integrated online condition monitoring system for nuclear power plants

International Nuclear Information System (INIS)

Hashemian, Hashem M.

2010-01-01

Online condition monitoring or online monitoring (OLM) uses data acquired while a nuclear power is operating to continuously assess the health of the plant's sensors, processes, and equipment; to measure the dynamic performance of the plant's process instrumentation; to verify in-situ the calibration of the process instrumentation; to detect blockages, voids, and leaks in the pressure sensing lines; to identify core flow anomalies; to extend the life of neutron detectors and other sensors; and to measure the vibration of reactor internals. Both the steady-state or DC output of plant sensors and their AC signal or noise output can be used to assess sensor health, depending on whether the application is monitoring a rapidly changing (e.g., core barrel motion) or a slowly changing (e.g., sensor calibration) process. The author has designed, developed, installed, and tested OLM systems (comprised of software/hardware-based data acquisition and data processing modules) that integrate low-frequency (1 mHz to 1 Hz) techniques such as RTD cross-calibration, pressure transmitter calibration monitoring, and equipment condition monitoring and high-frequency (1 Hz to 1 kHz) techniques such as the noise analysis technique. The author has demonstrated the noise analysis technique's effectiveness for measuring the dynamic response-time of pressure transmitters and their sensing lines; for performing predictive maintenance on reactor internals; for detecting core flow anomalies; and for extending neutron detector life. Integrated online condition monitoring systems can combine AC and DC data acquisition and signal processing techniques, can use data from existing process sensors during all modes of plant operation, including startup, normal operation, and shutdown; can be retrofitted into existing PWRs, BWRs, and other reactor types and will be integrated into next-generation plants. (orig.)

Integrated online condition monitoring system for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Hashemian, Hashem M. [Analysis and Measurement Services Corporation, Knoxville, TN (United States). AMS Technology Center

2010-09-15

Online condition monitoring or online monitoring (OLM) uses data acquired while a nuclear power is operating to continuously assess the health of the plant's sensors, processes, and equipment; to measure the dynamic performance of the plant's process instrumentation; to verify in-situ the calibration of the process instrumentation; to detect blockages, voids, and leaks in the pressure sensing lines; to identify core flow anomalies; to extend the life of neutron detectors and other sensors; and to measure the vibration of reactor internals. Both the steady-state or DC output of plant sensors and their AC signal or noise output can be used to assess sensor health, depending on whether the application is monitoring a rapidly changing (e.g., core barrel motion) or a slowly changing (e.g., sensor calibration) process. The author has designed, developed, installed, and tested OLM systems (comprised of software/hardware-based data acquisition and data processing modules) that integrate low-frequency (1 mHz to 1 Hz) techniques such as RTD cross-calibration, pressure transmitter calibration monitoring, and equipment condition monitoring and high-frequency (1 Hz to 1 kHz) techniques such as the noise analysis technique. The author has demonstrated the noise analysis technique's effectiveness for measuring the dynamic response-time of pressure transmitters and their sensing lines; for performing predictive maintenance on reactor internals; for detecting core flow anomalies; and for extending neutron detector life. Integrated online condition monitoring systems can combine AC and DC data acquisition and signal processing techniques, can use data from existing process sensors during all modes of plant operation, including startup, normal operation, and shutdown; can be retrofitted into existing PWRs, BWRs, and other reactor types and will be integrated into next-generation plants. (orig.)

Stability monitoring of a natural-circulation-cooled boiling water reactor

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der.

1989-01-01

Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

Life extension activities and modernization strategies for instrumentation ampersand control systems of research and power reactors in India

International Nuclear Information System (INIS)

Chaganty, S.P.; Bairi, B.R.

1993-01-01

Based on three and half decades of experience gained in the operation and maintenance of Instrumentation and Control Systems of nuclear reactors in India, specific investigations were made to understand various aspects of aging. The analysis of the failure rates of various instruments, plant outage figures and obsolescence of components have necessitated the replacement of instrumentation to improve the reliability and performance. The aging models available were used to determine the extent of performance degradation and to formulate maintenance strategies. The nuclear instrumentation of the aging research reactors at Bhabha Atomic Research Centre (BARC) has been replaced with high reliability equipment using modern integrated circuits. This has resulted in an improvement in the mean time between failure (MTBF) by a factor of five. The neutronic instrumentation of Fast Breeder Test Reactor (FBTR) at Madras is currently being upgraded with the introduction of microprocessor based safety units for reactivity computation and online testing of safety logic with Fine Impulse Technique. The operating experience has also indicated the necessity of developing online surveillance methods and status monitoring of various systems to detect aging. Online cable insulation measurement technique and noise analysis methods for vibration monitoring have been developed. Campbell method of signal processing has been successfully used in extending the useful life of Local Power Range monitors in the Boiling Water Reactor at Tarapur. In order to improve reliability, accuracy and provide efficient man machine interface, microprocessor based systems with online testing features have been installed in power reactors. These include the high performance reactor regulating system and centralised radiation monitoring systems commissioned at Kakrapara power station. The paper describes the above systems and the modernization strategies for nuclear instrumentation and control

Centralized environmental radiation monitoring system in JAERI

International Nuclear Information System (INIS)

Katagiri, H.; Kobalyashi, H.

1993-01-01

JAERI has continued the environmental radiation background survey and monitoring to ensure the safety of the peoples around the institute since one year before the first criticality of JRR-1 (Japan Research Reactor No.1) in August 1957. Air absorbed doses from β and γ radiation, α and β radioactivity in air and the radioactivities in environmental samples were the monitoring items. For the monitoring of β and γ radiation and α and β radioactivity in air, monitoring station and the centralized automatic environmental radiation monitoring system applying a computer were established as a new challenging monitoring system for nuclear facility, which was the first one not only in Japan but also in the would in 1960 and since then the system has been renewed two times (in 1973 and 1988) by introducing the latest technology in the fields of radiation detection and computer control at each stage. Present system renewed in 1988 was designed to prevent the interruption of monitoring due to computer troubles, communication troubles and power failures especially an instant voltage drop arisen from thunder by reflecting the experiences through the operation and maintenance of the former system. Dual telemeters whose power is constantly supplied via batteries (capable of 10 min monitoring after power failure) are equipped in the monitoring center to cope with telemeter troubles, which has operated successfully without any suspension being attributable to the power failures and telemeter troubles

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

International Nuclear Information System (INIS)

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

2013-06-01

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

REACTOR - a Concept for establishing a System-of-Systems

Science.gov (United States)

Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

2014-05-01

REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

Adapting a reactor safety assessment system for specific plants

International Nuclear Information System (INIS)

Ballard, T.L.; Cordes, G.A.

1991-01-01

The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

Neutron monitoring system

International Nuclear Information System (INIS)

Okido, Fumiyasu; Arita, Setsuo.

1994-01-01

The present invention concerns neutron monitoring for monitoring reactor power, and presents a generation state of abnormal signals by monitoring output signals from neutron sensors, judges abnormal signals at an excessively high level outputted from the sensors to a measuring operator or a reactor operator. That is, a threshold value judging means judges whether a sensor signal exceeds a predetermined threshold value or not. When it exceeds the value, recognition signals are outputted to a memory means. The memory means memorizes the times of input of the recognition signals on every period of interval signals outputted from a reference signal generation means. The memory content of the memory means and the previously inputted hysteresis of the sensor are compared and judged, to determine the extent of the degradation of the sensors and output the result of the judgement and hysteresis information to the display means. The input means accesses to the judging means and the memory means to retrieve and correct the content of the memory means and the hysteresis information inputted to the judging means. (I.S.)

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Sensor response monitoring in pressurized water reactors using time series modeling

International Nuclear Information System (INIS)

Upadhyaya, B.R.; Kerlin, T.W.

1978-01-01

Random data analysis in nuclear power reactors for purposes of process surveillance, pattern recognition and monitoring of temperature, pressure, flow and neutron sensors has gained increasing attention in view of their potential for helping to ensure safe plant operation. In this study, application of autoregressive moving-average (ARMA) time series modeling for monitoring temperature sensor response characteristrics is presented. The ARMA model is used to estimate the step and ramp response of the sensors and the related time constant and ramp delay time. The ARMA parameters are estimated by a two-stage algorithm in the spectral domain. Results of sensor testing for an operating pressurized water reactor are presented. 16 refs

Challenges for remote monitoring and control of small reactors

Energy Technology Data Exchange (ETDEWEB)

Trask, D., E-mail: traskd@aecl.ca [Atomic Energy of Canada Limited, Fredericton, New Brunswick (Canada)

2013-07-01

This paper considers a model for small, unmanned, remotely located reactors and discusses the ensuing cyber security and operational challenges for monitoring and control and how these challenges might be overcome through some of AECL's research initiatives and experience. (author)

Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

International Nuclear Information System (INIS)

Edwards, Robert M.

2003-01-01

The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs

Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

Science.gov (United States)

Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

2017-07-01

In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

Slow control systems of the Reactor Experiment for Neutrino Oscillation

International Nuclear Information System (INIS)

Choi, J.H.; Jang, H.I.; Choi, W.Q.; Choi, Y.; Jang, J.S.; Jeon, E.J.; Joo, K.K.; Kim, B.R.; Kim, H.S.; Kim, J.Y.; Kim, S.B.; Kim, S.Y.; Kim, W.; Kim, Y.D.; Ko, Y.J.; Lee, J.K.; Lim, I.T.; Pac, M.Y.; Park, I.G.; Park, J.S.

2016-01-01

The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, J.; Vocka, R.

2009-01-01

The SCORPIO-VVER system includes following features: 1) Validation of plant measurements and identification of sensor failures. 2) Optimum combination of measurements and calculations to obtain precise values of important parameters. 3) On-line 3D power distribution calculation with pin power reconstruction. 4) Limit checking and thermal margin calculation allowing for surveillance of VVER core limits such as DNBR, Sub-cooling margin, FdH and FQ peeking factors. 5) Integrated modules for monitoring fuel performance and coolant activity for identification of fuel failures. 6) Predictive capabilities and strategy planning, offering the possibility to check the consequences of operational manoeuvres in advance, prediction of critical parameters, etc. 7) Convenient monitoring of approach to criticality during reactor start-up. 8) Automated transition between cycles (fuel reload). The SCORPIO-VVER core monitoring system with its flexible and modular framework successfully responses to the plant operating needs and advances in nuclear fuel cycle strategies and fuel design. Modular framework allows for easy modifications of the system and implementation of new methods in physical modules. Even if the system is installed only on VVER-440 reactors, it could be adapted for VVER-1000 needs

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

Surveillance of nuclear power reactors

International Nuclear Information System (INIS)

Marini, J.

1983-01-01

Surveillance of nuclear power reactors is now a necessity imposed by such regulatory documents as USNRC Regulatory Guide 1.133. In addition to regulatory requirements, however, nuclear reactor surveillance offers plant operators significant economic advantages insofar as a single day's outage is very costly. The economic worth of a reactor surveillance system can be stated in terms of the improved plant availability provided through its capability to detect incidents before they occur and cause serious damage. Furthermore, the TMI accident has demonstrated the need for monitoring certain components to provide operators with clear information on their functional status. In response to the above considerations, Framatome has developed a line of products which includes: pressure vessel leakage detection systems, loose part detection systems, component vibration monitoring systems, and, crack detection and monitoring systems. Some of the surveillance systems developed by Framatome are described in this paper

I and C system at TRIGA - ICN reactor after more than 20 years of operation

International Nuclear Information System (INIS)

Ionila, M.; Preda, M.

2002-01-01

An I and C system that is involved in a nuclear safety function has to be itself safe in operation, strictly performing the survey of those parameters, which are linked to the safety function. The precision of such a system is sufficient for the safety function, but for a more accurate evaluation of the in-core experimental phenomena, the presence of a data acquisition and processing system is needed. The two systems must be together taken into account by the reactor operation. The data acquisition and processing system designated for the monitoring of the stationary or the slow-varying processes allow the safety function evaluation from the point of view of the statistics of the effective reactor operation time along a certain period of time. The evaluation of the unscheduled reactor shutdowns determined by those reactor systems having safety functions with the percentage contribution of each system is presented. The data were selected from the annual operation reports for the reactor and the reactor installations in the period 1981-1999

Reactor instrumentation and control

International Nuclear Information System (INIS)

Wach, D.; Beraha, D.

1980-01-01

The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

Nuclear reactor safety systems

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1980-01-01

A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

FFTF reactor assembly system technology

International Nuclear Information System (INIS)

Mangelsdorf, T.A.

1975-01-01

An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

System of ecological monitoring and assessment (EMA) for nuclear power plant site areas

International Nuclear Information System (INIS)

Vorob'ev, E.I.; Oleinikov, N.F.; Reznichenko, V.Yu.

1988-01-01

The goal of the EMA program is to improve the utilization of the natural and labor resources of the city of Sosnovyi Bor due to the opportune discovery of negative changes in the state of the surrounding environment and human health and to predict and assess the possible consequences of implementing technological and economic decisions and their optimization from the point of view of ecology with regard to reactor siting and sizing. The EMA system provides monitoring means and a data bank on the environment and health of workers and the population, a packet of programs for processing and analyzing data, and a set of models to describe and predict the condition of the object/environment/man system. The system monitors contamination sources, monitors and assesses aquatic and terrestrial ecosystems, and accounts for meteorological and biosolar factors which impact on reactor siting

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).

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

Computer measurement system of reactor period for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1997-01-01

The author simply introduces the hardware, principle, and software of the reactor period computer measure 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 computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

Fuel handling system of nuclear reactor plants

International Nuclear Information System (INIS)

Faulstich, D.L.

1991-01-01

This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

Development of CANDU core monitoring system

International Nuclear Information System (INIS)

Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H.

2003-01-01

The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1

Development of CANDU core monitoring system

Energy Technology Data Exchange (ETDEWEB)

Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H. [Institute for Advanced Engineering, Yongin (Korea, Republic of)

2003-07-01

The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1.

Reactor noise monitoring device

International Nuclear Information System (INIS)

Yamanaka, Hiroto.

1990-01-01

The present invention concerns a reactor noise monitoring device by detecting abnormal sounds in background noises. Vibration sounds detected by accelerometers are applied to a loose parts detector. The detector generates high alarm if there are sudden impact sounds in the background noises and applies output signals to an accumulation device. If there is slight impact sounds in the vicinity of any of the accelerometers, the accumulation device accumulates the abnormal sounds assumed to be generated from an identical site while synchronizing the waveforms for all of the channels. Then, the device outputs signals in which the background noises are cancelled, as detection signals. Therefore, S/N ratio can be improved and the abnormal sounds contained in the background noises can be detected, to thereby improve the accuracy for estimating the position where the abnormal sounds are generated. (I.S.)

Calibration method of radiation monitoring system at TQNPC

International Nuclear Information System (INIS)

Liu Zhengshan; Zhang Qingli; Liu Jinjin; Miao Yuxing; Geng Lixin; Zhuang Yun; Dong Jianfeng; He Change

2009-04-01

The calibration methods and calibration device for standard monitor of radioactive particulate, iodine, noble gas and so on are not yet set up at home. On consideration of the present situation of the radiation monitoring system at the Third Qinshan Nuclear Power Co. Ltd., we have studied the calibration method of these radiation monitoring instruments used for measuring the waste liquid, particulate, iodine and noble gas produced during the operation of nuclear reactor. Through the check against these instruments during the No. 202 and No. 103 overhaul, we got initially the method of the calibration and obtained the transfer coefficient of calibration when secondary solid sources are used for calibration. Through the testing and calibration, the credibility of the radiation monitoring system is enhanced. And at the same time, the problems existing in the calibration are discussed. (authors)

Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

2013-04-04

Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

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

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)

Auxiliary control system of the safety parameters for IPR-R1 reactor

International Nuclear Information System (INIS)

Coura, J.G.

1986-01-01

This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

Nuclear reactors. Use of the protection system for non-safety purposes (International Electrotechnical Commission Standard Publication 639:1979)

International Nuclear Information System (INIS)

Stefanik, J.

1996-01-01

This standard applies to the protection system of a nuclear reactor and, more especially, to all interconnections between a reactor protection system (as defined and explained in International Electrotechnical Commission Publication 231 A, first supplement to Publication 231, General Principles of Nuclear Reactor Instrumentation) and all other systems and equipment not part of the protection system, except: a) the physical connection between sensors of the protection system and the physical variables that they monitor, such as for example, thermo wells, moderating medium for neutron sensors, etc.; b) the electrical connection between the protection system and the reactor control rods or other safety mechanism; c) the electrical and pneumatic connections to the power distribution system (mains) and pneumatic supplies that supply power to the protection system. Although many clauses relate to all reactor protection systems, this standard applies mainly to protection systems in nuclear power reactors

In-operation diagnostic system for reactor coolant pump

International Nuclear Information System (INIS)

Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

1996-01-01

A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

Development of the temperature field at the WWER-440 core outlet monitoring system and application of the data analyses methods

International Nuclear Information System (INIS)

Spasova, V.; Georgieva, N.; Haralampieva, Tz.

2001-01-01

On-line internal reactor monitoring by 216 thermal couples, located at the reactor core outlet, is carried out during power operation of WWER-440 Units 1 and 2 at Kozloduy NPP. Automatic monitoring of technology process is performed by IB-500MA, which collects and performs initial data processing (discrediting and conversion of analogue signals into digital mode). The paper also presents the results and analyses of power distribution monitoring during the past 21-th and current 22-th fuel cycle at Kozloduy NPP, Unit 1 by using archiving system capacity and related software. The possibility to perform operational assessment and analysis of power distribution in the reactor core in each point of the fuel cycle is checked by comparison of the neutron-physical calculation results with reactor coolant system parameters. Paper shows that the processing and analysis of accumulated significant amount of data in the archive files increases accuracy and reliability of power distribution monitoring in the reactor core in each moment of the fuel cycle of WWER-440 reactors at Kozloduy NPP

Quantitative monitoring of an activated sludge reactor using on-line UV-visible and near-infrared spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Sarraguca, Mafalda C.; Lopes, Joao A. [Universidade do Porto, REQUIMTE, Servico de Quimica-Fisica, Faculdade de Farmacia, Porto (Portugal); Paulo, Ana; Alves, Madalena M.; Dias, Ana M.A.; Ferreira, Eugenio C. [Universidade do Minho, IBB - Institute for Biotechnology and Bioengineering, Centre of Biological Engineering, Braga (Portugal)

2009-10-15

The performance of an activated sludge reactor can be significantly enhanced through use of continuous and real-time process-state monitoring, which avoids the need to sample for off-line analysis and to use chemicals. Despite the complexity associated with wastewater treatment systems, spectroscopic methods coupled with chemometric tools have been shown to be powerful tools for bioprocess monitoring and control. Once implemented and optimized, these methods are fast, nondestructive, user friendly, and most importantly, they can be implemented in situ, permitting rapid inference of the process state at any moment. In this work, UV-visible and NIR spectroscopy were used to monitor an activated sludge reactor using in situ immersion probes connected to the respective analyzers by optical fibers. During the monitoring period, disturbances to the biological system were induced to test the ability of each spectroscopic method to detect the changes in the system. Calibration models based on partial least squares (PLS) regression were developed for three key process parameters, namely chemical oxygen demand (COD), nitrate concentration (N-NO{sub 3}{sup -}), and total suspended solids (TSS). For NIR, the best results were achieved for TSS, with a relative error of 14.1% and a correlation coefficient of 0.91. The UV-visible technique gave similar results for the three parameters: an error of {proportional_to}25% and correlation coefficients of {proportional_to}0.82 for COD and TSS and 0.87 for N-NO{sub 3}{sup -}. The results obtained demonstrate that both techniques are suitable for consideration as alternative methods for monitoring and controlling wastewater treatment processes, presenting clear advantages when compared with the reference methods for wastewater treatment process qualification. (orig.)

Quantitative monitoring of an activated sludge reactor using on-line UV-visible and near-infrared spectroscopy.

Science.gov (United States)

Sarraguça, Mafalda C; Paulo, Ana; Alves, Madalena M; Dias, Ana M A; Lopes, João A; Ferreira, Eugénio C

2009-10-01

The performance of an activated sludge reactor can be significantly enhanced through use of continuous and real-time process-state monitoring, which avoids the need to sample for off-line analysis and to use chemicals. Despite the complexity associated with wastewater treatment systems, spectroscopic methods coupled with chemometric tools have been shown to be powerful tools for bioprocess monitoring and control. Once implemented and optimized, these methods are fast, nondestructive, user friendly, and most importantly, they can be implemented in situ, permitting rapid inference of the process state at any moment. In this work, UV-visible and NIR spectroscopy were used to monitor an activated sludge reactor using in situ immersion probes connected to the respective analyzers by optical fibers. During the monitoring period, disturbances to the biological system were induced to test the ability of each spectroscopic method to detect the changes in the system. Calibration models based on partial least squares (PLS) regression were developed for three key process parameters, namely chemical oxygen demand (COD), nitrate concentration (N-NO(3)(-)), and total suspended solids (TSS). For NIR, the best results were achieved for TSS, with a relative error of 14.1% and a correlation coefficient of 0.91. The UV-visible technique gave similar results for the three parameters: an error of approximately 25% and correlation coefficients of approximately 0.82 for COD and TSS and 0.87 for N-NO(3)(-) . The results obtained demonstrate that both techniques are suitable for consideration as alternative methods for monitoring and controlling wastewater treatment processes, presenting clear advantages when compared with the reference methods for wastewater treatment process qualification.

WARMS - a continuous on-line environmental and emergency radiation monitoring system

International Nuclear Information System (INIS)

Ramsden, D.

1984-01-01

The Winfrith Airborne Release Monitoring System (WARMS) is used to monitor the environment around the Winfrith reactor site. It operates continuously monitoring the background radiation at 16 outstations and can provide rapid information should an accidental release occur. WARMS was developed jointly by the Radiological Safety Division and the Control and Instrumentation Division at Winfrith in association with the Safety and Reliability Directorate at Culcheth which developed the software. The system became operational in the autumn of 1983 and has since demonstrated a high degree of reliability and effectiveness. (author)

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

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

Data acquisition and signal processing system for IPR R1 TRIGA-Mark I nuclear research reactor of CDTN

International Nuclear Information System (INIS)

Mesquita, A.Z.; Maretti, F. Jr.; Rezende, H.C.; Tambourgi, E.B.

2004-01-01

The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. The main operational parameters were monitored by analog recorders and counters located in the reactor control console. The reactor operators registered the most important operational parameters and data in the reactor logbook. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients in some operations. A PC-based data acquisition was developed for the reactor that allows online monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that were not measured, like the power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measuring out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor; Desarrollo de un sistema de vigilancia remota, por medio del monitoreo de parametros claves de seguridad, para un reactor nuclear de investigacion

Energy Technology Data Exchange (ETDEWEB)

Urcia, Agustin; Arrieta, Rolando [Direccion de Produccion, Instituto Peruano de Energia Nuclear, Lima (Peru); Baltuano, Oscar; Chan, Renzo [Direccion de Investigacion y Desarrollo, Instituto Peruano de Energia Nuclear, Lima (Peru); Tincopa, Jean Pierre [Facultad de Ingenieria Electrica y Electronica, Universidad Nacional del Callao, Callao (Peru); Urquizo, Rafael [Facultad de Ingenieria Electronica, Universidad Tecnologica del Peru, Lima (Peru); Rosas, Bernick [Facultad de Ingenieria Electronica, Universidad Nacional de Ingenieria, Lima (Peru)

2014-07-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

Reactor protection and shut-down system

International Nuclear Information System (INIS)

Klar

1980-01-01

The reactor protection system being a part of the reactor safety system. The requirements on the reactor protection system are: high safety with regard to signal processing, high availability, self-reporting of faults etc. The functional sections of the reactor protection system are the analog section, the logic section and the generating of output signals. Description of the operation characteristics and of the extension of function. (orig.)

Monitoring device for the reactor pipelines

International Nuclear Information System (INIS)

Fukumoto, Akira.

1983-01-01

Purpose: To enable rapid and accurate operator's monitoring for the state of pipelines in a BWR type reactor. Constitution: Specific symbols are attached respectively to a fluid supply source constituting the pipelines of a nuclear reactor facility, a plurality of fluid passing points and equipments to be supplied with the fluid, and a symmetrical matrix comprising these symbols in rows and columns is constituted. Then, a matrix is prepared based on detection signals for the states of the liquid supply source, equipments to be supplied with fluid and pipeline equipments by rendering the matrix elements between the signals expressing the state capable of passing the fluid as 1 and the matrix elements between the signals expressing the state incapable of passing the fluid as 0 . The matrix thus prepared in a signal procession circuit and a matrix in a memory circuit previously storing the matrix expressing the normal state of the pipelines are compared to judge the state of the pipelines in a short time and with no misjudging. (Moriyama, K.)

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Development of an X Window based operator's interface for a core monitoring system

International Nuclear Information System (INIS)

Vegh, J.; Huszar, J.; Laz, J.

1992-09-01

The components, functioning and programming concepts of the man-machine interface applied in an upgraded version of the core monitoring system and reactor information system VERONA for WWER-440 type nuclear power reactors, installed at the Paks Nuclear Power Plant, are described. The application of the X Window standard Graphical User Interface facilitated modular interface design and made program development easier and faster. (author) 3 refs.; 13 figs

The micro-processor controlled process radiation monitoring system for reactor safety systems

International Nuclear Information System (INIS)

Mizuno, K.; Noguchi, A.; Kumagami, S.; Gotoh, Y.; Kumahara, T.; Arita, S.

1986-01-01

Digital computers are soon expected to be applied to various real-time safety and safety-related systems in nuclear power plants. Hitachi is now engaged in the development of a micro-processor controlled process radiation monitoring system, which operates on digital processing methods employed with a log ratemeter. A newly defined methodology of design and test procedures is being applied as a means of software program verification for these safety systems. Recently implemented micro-processor technology will help to achieve an advanced man-machine interface and highly reliable performance. (author)

A method for on-line reactivity monitoring in nuclear reactors

International Nuclear Information System (INIS)

Dulla, S.; Nervo, M.; Ravetto, P.

2014-01-01

Highlights: • The problem of the on-line monitoring of reactivity in a source-free nuclear reactor is considered. • A relationship between the system stable period and the power, its derivative and its integral is derived. • The reactivity can be reconstructed at each time instant from the measured power-related quantities. • A study on the sensitivity of the reactivity to the uncertainty on the values of the integral parameters is performed. • The spatial effects are investigated by applying the method to the interpretation of flux signals. - Abstract: In the present work the problem of the on-line monitoring of the reactivity in a source-free nuclear reactor is considered. The method is based on the classic point kinetic model of reactor physics. A relationship between the instantaneous value of the system stable period and the values of the neutron flux amplitude (or the power), of its derivative and of the integral convolution term determining the instantaneous value of the effective delayed neutron concentration is derived. The reactivity can then be evaluated through the application of the inhour equation, assuming the effective delayed neutron fraction and prompt generation time are known from independent measurements. Since the power related quantities can be assumed to be experimental observables at each instant, the reactivity can be easily reconstructed. The method is tested at first through the interpretation of power histories simulated by the solution of the point kinetic equations; the effect of the time interval between power detections on the accuracy is studied, proving the excellent performance of the procedure. The work includes also a study on the sensitivity of the reactivity forecast to the uncertainty on the values of the effective delayed neutron fraction and prompt generation time. The spatial effects are investigated by applying the method to the interpretation of flux evolution histories generated by a numerical code solving

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

DIMOS: a new generation of nuclear power plant process monitoring systems