ISAT promises fail-safe computer-based reactor protection

International Nuclear Information System (INIS)

Anon.

1989-01-01

AEA Technology's ISAT system is a multiplexed microprocessor-based reactor protection system which has very extensive self-monitoring capabilities and is inherently fail safe. It provides a way of addressing software reliability problems that have tended to hamper widespread introduction of computer-based reactor protection. (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.)

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

New measuring and protection system at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Jurickova, M.

2006-01-01

The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

RA Research nuclear reactor, Part II: radiation protection at the RA reactor in 1987

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1987-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1988-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

Validation of reactor core protection system

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Additional reactor protection system of RBMK-1500

International Nuclear Information System (INIS)

1999-01-01

Analysis of anticipated transients without scram of RBMK-1500 reactor showed that additional reactor protection system is required. Data of accident analysis in the case of loose of external electric power and loose of vacuum in condensers of turbines are provided

Safety Analysis for Power Reactor Protection System

International Nuclear Information System (INIS)

Eisawy, E.A.; Sallam, H.

2012-01-01

The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

Physical protection of power reactors

International Nuclear Information System (INIS)

Darby, J.L.

1979-01-01

Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant

Design and testing of integrated circuits for reactor protection channels

International Nuclear Information System (INIS)

Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

1995-01-01

Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built

Nuclear Reactor RA Safety Report, Vol. 9, Radiation protection

International Nuclear Information System (INIS)

1986-11-01

Instrumentation for Radiation protection existing at the RA reactor is dating mostly from the period 1957-1959 when the reactor has been built. With some minor exception it was produced in USSR. Radiation protection system was constructed based on specific design project, somewhat modified original USSR project which has been indispensable because of some modification of the building design. During the past 27 years no renewal of the instrumentation was done, only maintenance was performed. Instrumentation consists of old electronic devices which caused difficulties and even prevented regular maintenance because of lack of spare parts. Instrumentation for radiation protection at the RA reactor is classified as follows: centralized dosimetry system; stationary dosimetry instrumentation, movable and personal dosimetry systems. Apart from the scheme of dosimetry instrumentation this volume includes description of radiation protection procedures; protection devices; radiation doses and dose limit data; program for environmental radioactivity control; medical control procedures [sr

A protection system of low temperature thermo-supply nuclear reactor

International Nuclear Information System (INIS)

Jiang Binsen

1988-09-01

A Protection system of low temperature thermo-supply nuclear reactor is introduced. It is the first protection system, which is designed and manufactred on the basis of Chinese National Standard GB 4083-83 'General Safety Principle of Nuclear Reactor Protection System', to be considered under the circumstances of industry level in China. Advantages of the protection system are as follows: 1)The single failure criteria can fully be fulfilled by the protection system. 2) On-line testing system can be used for detecting all of failure components and quick identifying the failure points in the system. 3) It is convenience for maintenacnce of the system. To complete this project is very important and helpful in promoting the development of the protection system and safety operation of nuclear reactor in China

Fault-tolerant reactor protection system

International Nuclear Information System (INIS)

Gaubatz, D.C.

1997-01-01

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. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs

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

Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

Lightning protection system analysis at Multipurpose Reactor G A. Siwabessy building

International Nuclear Information System (INIS)

Teguh-Sulistyo

2003-01-01

Analysis to the part of lightning protection system at Multi Purpose Reactor GA Siwabessy (RSG-GAS) have been done. Observation examined the damage of some part of the earthing system caused by human error of chemically system. The analysis performed some assumptions and simulations to the points of lightning stroke. From this analysis obtained that the reactor building do not have vertical finial which can protect effectively to the whole reactor building and auxiliary building. Installing some new finials at some places are needed to protect building therefore the reactor building and auxiliary building well safe from lighting stroke

Radiation protection at new reactors

International Nuclear Information System (INIS)

Brissaud, A.

2000-01-01

The theoretical knowledge and the feedback of operating experience concerning radiations in reactors is now considerable. It is available to the designer in the form of predictive softwares and data bases. Thus, it is possible to include the radiation protection component throughout all the design process. In France, the existing reactors have not been designed with quantified radiation protection targets, although considerable efforts have been made to reduce sources of radiation illustrated by the decrease of the average dose rates (typically a factor 5 between the first 900 MWe and the last 1300 MWe units). The EDF ALARA PROJECT has demonstrated that good practises, radiation protection awareness, careful work organization had a strong impact on operation and maintenance work volume. A decrease of the average collective dose by a factor 2 has been achieved without noticeable modifications of the units. In the case of new nuclear facilities projects (reactor, intermediate storage facility,...), or special operations (such as steam generator replacement), quantified radiation protection targets are included in terms of collective and average individual doses within the frame of a general optimization scheme. The target values by themselves are less important than the application of an optimization process throughout the design. This is because the optimization process requires to address all the components of the dose, particularly the work volume for operation and maintenance. A careful study of this parameter contributes to the economy of the project (suppression of unecessary tasks, time-saving ergonomy of work sites). This optimization process is currently applied to the design of the EPR. General radiation protection provisions have been addressed during the basic design phase by applying general rules aiming at the reduction of sources and dose rates. The basic design optimization phase has mainly dealt with the possibility to access the containment at full

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Science.gov (United States)

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... perform their duties. (6) Prior to entry into a material access area, packages shall be searched for...

Design and testing of integrated circuits for reactor protection channels

International Nuclear Information System (INIS)

Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

1995-01-01

Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing

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)

Development of digital power measuring and protecting equipment for SPRR-300 reactor

International Nuclear Information System (INIS)

Wang Xuejie; Li Xi'an; Zhu Shilei

2005-01-01

A measuring and protecting equipment of reactor power based on Single-Chip Microcomputer is introduced in this paper. The composition of hardware and the major control idea about the software for the equipment are presented. Digitizing the measuring data from nuclear instruments is precondition of reactor control and protection system which would be computerizing, and it is also an application of redundancy and variety of reactor protection system in nuclear measuring instruments. At last the working state is described. (authors)

Franco-German cooperation for the physical protection of the EPR reactor

International Nuclear Information System (INIS)

Jalouneix, J.; Hagemann, A.

2001-01-01

This article presents the proceeding that has been followed in the EPR (European pressurized water reactor) project concerning physical protection against malevolent actions and robbery of nuclear materials. Before the different options of the nuclear island were definitely set, a task group had been constituted to examine if these options could hamper the setting of physical protection measures that are required by the legislation of the 2 countries. Another group composed of experts from IPSN/GRS (Institut de Protection et de Surete Nucleaire / Gesellschaft fur Anlagen und Reaktorsicherheit) had the task to define common requirements concerning the physical protection of reactors in Germany and in France. In this framework the EPR project team has prepared a technical document reviewing the different dispositions that have been retained to assure the physical protection of the reactor. (A.C.)

Radiation protection issues for EPR reactor

International Nuclear Information System (INIS)

Miniere, D.; Le Guen, B.; Beneteau, Y.; Le Guen, B.

2008-01-01

As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection Radiation Protection aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing Radiation Protection-significant activities has been to include all involved stakeholders - designers, licensee and contractor companies - in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The Radiation Protection target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 Man Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the E.P.R.'s reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), Radiation Protection studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the

Radiation protection in a university TRIGA reactor

International Nuclear Information System (INIS)

Tschurlovits, M. . Author

2004-01-01

Radiation protection in a university institute operating a research reactor and other installations has different constraints as a larger facility. This is because the legal requirements apply in full, but the potential of exposure is low, and accesses has to be made available for students, but also for temporary workers. Some of the problems in practical radiation protection are addressed and solutions are discussed. In addition, experience with national radiation protection legislation recently to be issued is addressed and discussed. (author)

Preventive protection device and method for bottom of reactor pressure vessel

International Nuclear Information System (INIS)

Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

1995-01-01

In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

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

Assessment of core protection and monitoring systems for an advanced reactor SMART

International Nuclear Information System (INIS)

In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

2002-01-01

Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

Automated testing of reactor protection instrumentation made easy

International Nuclear Information System (INIS)

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

1997-01-01

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

Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

1998-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

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

Radiation Protection at Light Water Reactors

CERN Document Server

Prince, Robert

2012-01-01

This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

International Nuclear Information System (INIS)

Rimpault, G.

2004-01-01

Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

Science.gov (United States)

Blanchet, David; Fontaine, Bruno

2017-09-01

The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

Software reliability and safety in nuclear reactor protection systems

Energy Technology Data Exchange (ETDEWEB)

Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

1993-11-01

Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

Software reliability and safety in nuclear reactor protection systems

International Nuclear Information System (INIS)

Lawrence, J.D.

1993-11-01

Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor

Automated reactor protection testing saves time and avoids errors

International Nuclear Information System (INIS)

Raimondo, E.

1990-01-01

When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

Comparison of radioactive doses after the last protection layer insight the reactor structure for Russian VVER-1000 and German PWR-1300 reactors

International Nuclear Information System (INIS)

Rahimi, A.; Mansourshaiflu, N.; Alizadeh, M. R.

2004-01-01

In pressurized reactors (VVER and PWR), various protections layers are used for reducing the output core doses. At any protection layer, some amount of neutron and gamma doses is reduced. In this project the axial flux of neutron and gamma beams have been evaluated at various protection layers in the operation state the German PWR-1300 and Russian VVER-1000 reactors by the MCNP computer code. For the purpose of effective use of the MCNP code and assuring its correct performance about of fluxed beams common and series of scientific answers and bench marks should be considered and the results obtained by the MCNP code, be compared with this answers. Then by using appropriate method, for reducing the flux variants of neutron and gamma beams at various protection layers of German PWR-1300 and Russian VVER-1000 reactors of the operation state of both reactors have been accelerated. In this projects, bench marks are computations and numbers existing in PSAR's present at Bushehr nuclear power plant. At the end, by using the results obtained and the standard doses, the time which a person can have work activity at the reactor wall (after the last protection layer), was compared for the operation status of the German PWR-1300 and Russian VVER-1000 reactors

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

Radiation protection personnel training in Research Reactors

International Nuclear Information System (INIS)

Fernandez, Carlos Dario; Lorenzo, Nestor Pedro de

1996-01-01

The RA-6 research reactor is considering the main laboratory in the training of different groups related with radiological protection. The methodology applied to several courses over 15 years of experience is shown in this work. The reactor is also involved in the construction, design, start-up and sell of different installation outside Argentina for this reason several theoretical and practical courses had been developed. The acquired experience obtained is shown in this paper and the main purpose is to show the requirements to be taken into account for every group (subjects, goals, on-job training, etc) (author)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Device for protecting deformations of reactor cores

International Nuclear Information System (INIS)

Kato, Yasuyoshi; Urushihara, Hiroshi.

1975-01-01

Object: To provide a fluid pressure cylinder, which is operated according to change in temperature of coolant for a reactor to restrain or release a core, to simply and effectively protect deformation of the core. Structure: A closed fluid pressure cylinder interiorly filled with suitable fluid is disposed in peripherally equally spaced relation in an annular space between a core barrel of a reactor and a reactor vessel. A piston is mounted in fluid-tight fashion in a plurality of piston openings made in the cylinder, the piston being slidably moved according to expansion and contraction of the fluid filled in the cylinder. The piston has a movable frame mounted at the foremost end thereof, the movable frame being moved integral with the piston, and the surface opposite the mount thereof biasing the outermost peripheral surface of the core. (Kamimura, M.)

Protection of semiconductor converters for controlled bypass reactors

International Nuclear Information System (INIS)

Dolgopolov, A. G.; Akhmetzhanov, N. G.; Karmanov, V. F.

2010-01-01

Possible ways of protecting thyristor converters in systems for magnetizing 110 - 500 kV controlled bypass reactors during switching and automatic reclosing are examined based on experience with the development of equipment, line tests, and mathematical modelling.

Reactor protection system including engineered features actuation system

International Nuclear Information System (INIS)

Palmaers, W.

1982-01-01

The safety concept requires to ensure that - the reactor protection system - the active engineered safeguard - and the necessary auxiliary systems are so designed and interfaced in respect of design and mode of action that, in the event of single component failure reliable control of the consequences of accidents remains ensured at all times and that the availability of the power plant is not limited unnecessarily. In order to satisfy these requirements due, importance was attached to a consistent spacial separation of the mutually redundant subsystems of the active safety equipment. The design and layout of the reactor protection system, of the power supply (emergency power supply), and of the auxiliary systems important from the safety engineering point of view, are such that their subsystems also largely satisfy the requirements of independence and spacial separation. (orig./RW)

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

A relay rack for a control and protection system for nuclear reactors

International Nuclear Information System (INIS)

Miyata, Yasuyuki; Oda, Noriaki; Akiyama, Toyoshi

1975-01-01

It is obvious that all the equipment in the various systems that constitute a nuclear power plant must exhibit the highest levels of reliability, but the reactor control and protection system is of vital importance, and thus it requires a particularly thorough approach, incorporating redundancy, independence and separation. The paper describes the functions, construction and specifications of the relay rack - one of the most important items of equipment for reactor control and protection in a generating facility using a pressurized-water reactor - and it gives details of the extent to which these three requirements are satisfied. (author)

Improving 200 MW NDHR reactor protection system with GAL devices

International Nuclear Information System (INIS)

Shi Mingde; Li Duo; Xie Zhengguo

1999-01-01

The emergence of General Array Logic (GAL), a fairly new type of logic devices with the characteristics of user-definable logic functions, have led to a revolutionary change in the design of logical circuits. The improvements of the reactor protection system for the 200 MW nuclear district heating reactor (NDHR) using GAL are covered

New technology for reactor protection system of CAREM reactor

International Nuclear Information System (INIS)

Dezzutti, J.C.; Verrastro, C.; Estryk, D.

2009-01-01

The use of FPGA in safety functions in a nuclear power plant, increase the reliability of software based systems, without loose any of the function required by the supervision and control systems. In this work the architecture of a Reactor Protection System is described, it use four independent measurement channels in 2 oo 4 configuration, each channel is based on diverse approach in 1 oo 2 configuration, the reliability of this system is near the same than the hardwired logic, with full performance like software based system. (author)

78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

Science.gov (United States)

2013-08-19

... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

Model Based Cyber Security Analysis for Research Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Model Based Cyber Security Analysis for Research Reactor Protection System

International Nuclear Information System (INIS)

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

2013-01-01

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

Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

International Nuclear Information System (INIS)

Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

2012-01-01

The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

System of nuclear power reactor protection using dynamic logic

International Nuclear Information System (INIS)

Carvalho, P.V.R. de; Silva, L.C.R.P. da

1990-01-01

The aim of this work is the design of a Reactor Protection System (RPS) using dynamic logic as basic circuitry principle. This concept was developed to permit the electronic and eletromagnetic components employment in 'fail-safe' mode applied to automatic shutdown systems. 'Fail-safe' here means that a fail always yields a constant state that leads to a plant shutdown condition. So the normal condition of operation corresponds to an oscillating state response and the fail or abnormal condition to a static one. At present, almost all modern nuclear plant reactor protection systems use dynamic logic, just differing in the kind of technology employed in the construction of the system. In this work we define what technology best fits our necessities, setting out to design a RPS based on this philosophy. (author) [pt

Evaluation of liquid metal protection of a limiter/divertor in fusion reactors

International Nuclear Information System (INIS)

Hassanein, A.M.; Smith, D.L.

1988-01-01

The liquid metal protection concept is proposed mainly to prolong the lifetime of a divertor or a limiter in a fusion reactor. This attractive idea for protection requires studying a wide range of problems associated with the use of liquid-metals in fusion reactors. In this work the protection by liquid-metals has concentrated on predictions of the loss rate of the film to the plasma, the operating surface temperatures required for the film, and the potential tritium inventory requirement. The effect of plasma disruptions on the liquid metal film is also evaluated. Other problems such as liquid metal compatibility with structural materials, magnetic field effects, and the effect of liquid metal contamination on plasma performance are discussed. Three candidate liquid-metals are evaluated, i.e., lithium, gallium, and tin. A wide range of reactor operating conditions valid for both near term machines (INTOR and ITER) and for the next generation commercial reactors (TPSS) are considered. This study has indicated that the evaporation rate for candidate liquid metals can be kept below the sputtering range for reasonable operating temperatures and plasma edge conditions. At higher temperatures, evaporation dominates the losses. Impurity transport calculations indicate that impurities from the plate should not reach the main plasma. One or two millimeters of liquid films can protect the structure from severe plasma disruptions. Depending on the design of the liquid metal protection system, the tritium inventory in the liquid film is predicted to be on the order of a few grams. 16 refs., 5 figs

Unavailability Analysis of the Reactor Core Protection System using Reliability Block Diagram

International Nuclear Information System (INIS)

Shin, Hyun Kook; Kim, Sung Ho; Choi, Woong Suk; Kim, Jae Hack

2006-01-01

The reactor core of nuclear power plants needs to be monitored for the early detection of core abnormal conditions to protect plants from a severe accident. The core protection calculator system (CPCS) has been provided to calculate the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) based on measured parameters of reactor and coolant system. The original CPCS for OPR 1000 has been designed and implemented based on the concurrent 3205 computer system whose components are obsolete. The CPCS based on Westinghouse Common-Q system has recently been implemented for the Shin-Kori Nuclear Power Plant, Units 1 and 2(SKN 1 and 2). An R and D project has been launched to develop new core protection system called as RCOPS (Reactor Core Protection System) with the partnership of KOPEC and Doosan Heavy Industries and Construction Co. RCOPS is implemented on the HFC-6000 safety class programmable logic controller (PLC). In this paper, the reliability of RCOPS is analyzed using the reliability block diagram (RBD) method. The calculated results are compared with that of the CPCS for SKN 1 and 2

The use of process computers in reactor protection systems

International Nuclear Information System (INIS)

1973-04-01

The report contains the papers presented at the LRA information meeting in spring 1972, concerning the use of process computers in reactor protection systems. The main interest was directed at a system conception as proposed from AEG for future BWR-plants. (orig.) [de

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)

Reactor protecting device

International Nuclear Information System (INIS)

Ono, Hiroshi; Kasuga, Hajime; Kasuga, Hiroshi.

1984-01-01

Purpose: To reduce the recycling flowrate thereby decrease the neutron flux level before the reactor shutdown upon generation of abnormality such as increase in the neutron flux, by setting the safety level lower than the value for generating the reaction scram signal. Constitution: A netron flux safety level setter and an instruction signal generator are disposed between a neutron flux detector and a recycling flowrate control device. A neutron flux safety level lower than the level for generating a reactor scram signal and higher that the level for the ordinary operation is set and, if the detection level for the neutron flux in the reactor core arrives at the safety level, a neutron flux decreasing instruction signal is outputted from the instruction signal generator to the recycling flowrate control device to thereby decrease the recycling flowrate and decrease the neutron flux without reaching the reactor shutdown, whereby the thermal safety of the fuel rod can be maintained and the reactor operation performance can be improved. (Moriyama, K.)

Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

International Nuclear Information System (INIS)

Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

2011-01-01

The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

Emergency systems and protection equipment of modular steam generators for fast reactors

International Nuclear Information System (INIS)

Matal, O.

The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

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

International Nuclear Information System (INIS)

Martins, Roque Hudson da Silva

2016-01-01

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

Qualification issues for advanced light-water reactor protection systems

International Nuclear Information System (INIS)

Korsah, K.; Clark, R.L.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

International Nuclear Information System (INIS)

Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

2012-01-01

Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

A structure for the protection of nuclear-reactor pressurized-vessels against rupture

International Nuclear Information System (INIS)

Marcellin, J.-P.; Aubert, Gilles

1974-01-01

Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

Twenty years of Radiology in RP-10 nuclear reactor protection

International Nuclear Information System (INIS)

Zapata, Alejandro L.; Ramos, Fernando T.; Arrieta, Rolando W.B.; Vela Mora, Mariano

2013-01-01

In this report we present the results about radiation controls during 1990 - 2010, carried out in the Nuclear Reactor RP-10 of the Nuclear Center of Huarangal. These controls and radiological evaluation are of much utility for the radio personnel protection of this one and other reactors, since it allows to compares these variables with respect to the time. From the results obtained in monitoring and radiation controls, we conclude that in no case it has been reached the limits allowed by the Peruvian Regulating Authority. (author)

New digital control and power protection system of VR 1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Matejka, K.; Juoeickova, M.

2005-01-01

The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

Development and Reliability Analysis of HTR-PM Reactor Protection System

International Nuclear Information System (INIS)

Li Duo; Guo Chao; Xiong Huasheng

2014-01-01

High Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM) digital Reactor Protection System (RPS) is a dedicated system, which is designed and developed according to HTR-PM NPP protection specifications. To decrease the probability of accident trips and increase the system reliability, HTR-PM RPS has such features as a framework of four redundant channels, two diverse sub-systems in each channel, and two level two-out-of-four logic voters. Reliability analysis of HTR-PM RPS is based on fault tree model. A fault tree is built based on HTR-PM RPS Failure Modes and Effects Analysis (FMEA), and special analysis is focused on the sub-tree of redundant channel ''2-out-of-4'' logic and the fault tree under one channel is bypassed. The qualitative analysis of fault tree, such as RPS weakness according to minimal cut sets, is summarized in the paper. (author)

Trip setpoint analysis for the reactor protection system of an advanced integral reactor

International Nuclear Information System (INIS)

Yang, Soo Hyung; Kim, Soo Hyung; Chung, Young Jong; Zee, Sung Quun

2007-01-01

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria

Reactor safety protection system

International Nuclear Information System (INIS)

Nishi, Hiroshi; Yokoyama, Tsuguo.

1989-01-01

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

Design of first reactor protection system prototype for C A R E M reactor

International Nuclear Information System (INIS)

Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

2006-01-01

In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Science.gov (United States)

2010-01-01

... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

Reactor protection system refurbishment at Paks

International Nuclear Information System (INIS)

Hetzmann, A.; Turi, T.

1997-01-01

The history and the milestones of the reactor protection system refurbishment are outlined. During the preparation phase of the refurbishment project, detailed requirements have been set up and specific technical solutions developed. The structure of the project documents prepared during these activities is shown in a figure. The life cycle of the project was divided into four phases: the preparatory phase; the design and manufacturing phase; the installation and commissioning phase; and the operation phase. For all four Paks units a time schedule for implementation was set up. The licensing process is dealt with; the principal license was issued in June 1996. (A.K.)

Criteria of the efficiency for radiation protection of tokamak reactor superconducting magnet coils

International Nuclear Information System (INIS)

Zimin, S.A.

1988-01-01

Factors determining serviceability of the main elements (superconductor, stabilizing conductor, insulation) of superconducting magnet coils for tokamak reactors are discussed. It is suggested that the limiting values of total and specific energy release in the material of superconducting coils, increase in electric resistance of the stabilizing conductor, decrease in the superconductor critical current and damage of the superconducting magnet insulation should be used as criteria of the reactor internal radiation protection efficiency. The conclusion is made that neutron fluence in the magnet coil components considered can be used as a generalized criterion of the first approximation for the evaluation of the protection efficiency

Burst protected nuclear reactor plant with PWR

International Nuclear Information System (INIS)

Harand, E.; Michel, E.

1978-01-01

In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

Method and practice on safety software verification and validation for digital reactor protection system

International Nuclear Information System (INIS)

Li Duo; Zhang Liangju; Feng Junting

2010-01-01

The key issue arising from digitalization of reactor protection system for Nuclear Power Plant (NPP) is in essence, how to carry out Verification and Validation (V and V), to demonstrate and confirm the software is reliable enough to perform reactor safety functions. Among others the most important activity of software V and V process is unit testing. This paper discusses the basic concepts on safety software V and V and the appropriate technique for software unit testing, focusing on such aspects as how to ensure test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed herein was successfully used in the work of unit testing on safety software of a digital reactor protection system. (author)

Radiation protection planning for decommissioning of research reactor facilities

International Nuclear Information System (INIS)

Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

2008-01-01

The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

Computer based systems for fast reactor core temperature monitoring and protection

International Nuclear Information System (INIS)

Wall, D.N.

1991-01-01

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

Study of Physical Protection System at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Ligam, A.S.; Ina, I.; Zarina Masood

2016-01-01

Physical protection program at PUSPATI TRIGA Reactor (RTP) which is located at Nuklear Malaysia, Bangi Complex has been strengthened and upgraded from time to time to accommodate current situation needs. However, there is always room for improvement. Hence, study have been made to look deeper into physical protection components such as delay systems, external sensors, PPS intruder alarm sensors, use of video system, personnel security or insider threats, access control operation system operation rules and security culture that may need to take into consideration. (author)

Comparison, with regard to safety, between a hard-wired reactor protection system and a computerized protection system. Pt. 1

International Nuclear Information System (INIS)

Buettner, W.E.

1976-07-01

The study compares a conventional hard-wired dynamic reactor protection system with a computerized protection system. In the comparison, only the unequivocally safety-oriented protection actions are considered. In the first part, the different structures of both systems and the method of verification for their functional safety will be described. In the second part, the mean unavailability in case of demand for both systems under defined conditions will be determined. (orig.) [de

Reliability analysis of the reactor protection system with fault diagnosis

International Nuclear Information System (INIS)

Lee, D.Y.; Han, J.B.; Lyou, J.

2004-01-01

The main function of a reactor protection system (RPS) is to maintain the reactor core integrity and reactor coolant system pressure boundary. The RPS consists of the 2-out-of-m redundant architecture to assure a reliable operation. The system reliability of the RPS is a very important factor for the probability safety assessment (PSA) evaluation in the nuclear field. To evaluate the system failure rate of the k-out-of-m redundant system is not so easy with the deterministic method. In this paper, the reliability analysis method using the binomial process is suggested to calculate the failure rate of the RPS system with a fault diagnosis function. The suggested method is compared with the result of the Markov process to verify the validation of the suggested method, and applied to the several kinds of RPS architectures for a comparative evaluation of the reliability. (orig.)

Reliability of redundant structures of nuclear reactor protection systems

International Nuclear Information System (INIS)

Vojnovic, B.

1983-01-01

In this paper, reliability of various redundant structures of PWR protection systems has been analysed. Structures of reactor tip systems as well as the systems for activation of safety devices have been presented. In all those systems redundancy is achieved by means of so called majority voting logic ('r out of n' structures). Different redundant devices have been compared, concerning probability of occurrence of safe as well as unsafe failures. (author)

Considerations for tritium protection at a fusion reactor

International Nuclear Information System (INIS)

Easterly, C.E.

1981-01-01

The purpose of this paper is to indicate the general direction of current fusion reactor concepts regarding tritium, and to indicate some options in tritium control strategies. Certain strategies, in addition to providing reduced potential health hazard, afford the potential for engineering alternatives for in-plant tritium control systems. The overall coupling of containment cleanup systems and health protection must continue to develop with increased knowledge of the health effects of different tritium species and the consequent systems options available subsequent to this understanding

75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

Science.gov (United States)

2010-10-13

... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The... nuclear fuel in transit? H. Why require a telemetric position monitoring system or an alternative tracking... nuclear fuel in transit. The interim final rule added 10 CFR 73.37, ``Requirements for Physical Protection...

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

El-Kafas, A.A.

1997-01-01

The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

Failsafe design criteria for computer based reactor protection systems

International Nuclear Information System (INIS)

Keats, A.B.

1980-01-01

The design criteria proposed are an extrapolation of the failsafe mode of operation used in the UK in hardwired reactor protection systems. This is achieved by making the operational condition of the reactor dependent upon an energetic state of the protection system components. An important objective of the proposed design criteria is to eliminate, or at least to minimize, the need for a failure-mode-and-effect-analysis (FMEA) of the computer software. This demands some well defined but simple constraints upon the way in which data are stored in the computers, but the objective is achieved almost entirely by hardware properties of the system. The first of these is the systematic use of hardwired test inputs which cause transient excursions into the tripped state in a uniquely ordered but easily recognizable sequence. The second is the use of hardwired pattern recognition logic which generates a dynamic healthy stimulus for the shutdown actuators only in response to the unique sequence formed by the hardwired input signal pattern. It therefore detects abnormal states of any of the system inputs, software errors, wiring errors and hardware failures. This hardwired logic is conceptually simple, failsafe, and is amenable to simple FMEA. (U.K.)

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

International Nuclear Information System (INIS)

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

2014-10-01

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

Reactor control and protection of full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Zhu Jinping; Sun Jiliang

1996-01-01

The control and protection simulation of Qinshan 300 MW Nuclear Power Unit, including the nuclear control, the pressurizer pressure control, the pressurizer level control, the rod control, the reactor shutdown protection and engineered safety feature etc are briefly introduced

A reactor core/containment status evaluation flowchart for determining protective actions in emergencies

International Nuclear Information System (INIS)

Glissman, M.A.

1988-01-01

In the event of an emergency at a power reactor station, there might not be adequate time or sufficient data to fully assess radiological implications and make protective action recommendations based on projected population exposures. Thus, decision-making guidance is needed that is based on readily available plant indicators, not just on time-consuming dose calculations. In the United States, this guidance must be compatible with the recommended by the Nuclear Regulatory Commission and the Environmental Protection Agency, and it must include predetermined, measurable, site-specific parameters for assessing conditions in the reactor core and containment. The preparation of this real time guidance calls for the selection of suitable parameters and the determination of the values for these parameters that will correspond to different levels of protective action. This process is illustrated in this paper by selecting parameters and determining appropriate values for constructing a Core/Containment Status Evaluation Flowchart for an example power plant

Environmental protection problems from the standpoint of regeneration of fast neutron reactor fuel

International Nuclear Information System (INIS)

Gedeonov, L.I.; Lazarev, L.N.; Suprunenko, A.N.

The discussion of the problem of environmental protection is based on two principles: a strict observance of legislatively established standards for permissible concentrations of radionuclides in objects of the environment and for dose loads for the population; all possible steps to reduce the contamination to a level justified in practice. Environmental protection steps are considered from the points of view of a systematic analysis. A survey of the environmental protection system near sources of radioactive discharges is given. The basic interactions and feedbacks are indicated. Characteristics differentiating the discharges of the fuel cycle of fast neutron breeder reactors from discharges of the slow neutron cycle are discussed. It is shown that it is necessary to study the overall regional and global interactions of discharges of the atomic power industry. The characteristics of situations at nuclear fuel cycle facilities of fast neutron reactors are discussed. The necessity of additional technical steps to prevent accidents and eliminate their effects if they take place is emphasized

Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations. Published on September 24, 2012

International Nuclear Information System (INIS)

Couturier, Jean; Bruna, Giovanni; Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Hache, Georges

2012-01-01

The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

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

International Nuclear Information System (INIS)

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

2003-03-01

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

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

EL-Kafas, A.E.A.E.

1996-01-01

the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

System for step-wise accident protection of nuclear reactors

International Nuclear Information System (INIS)

Rubek, J.; Kuklik, B.; Bednarik, K.

1991-01-01

A system comprising electric switching circuits is proposed for the control of a WWER type reactor shutdown in case of turbine failure or another abnormal situation. The fastest reactor shutdown mode is only resorted to if the pressures in the primary and secondary circuits would otherwise increase above tolerable limits and safety valves would open. The temperature and pressure stress of the nuclear power plant components and fuel is reduced. In this manner, the losses emerging during turbine failures due to false alarms are minimized. The contacts of the system switch if the turbines are relieved to the power of the unit home consumption, if the first or second turbine fails by closing the quick-acting valves, if a signal for blocking the by-pass stations of the operated turbines appears, or if the electric supply of the control system and of the turbo-set protection fails. (M.D.). 1 fig

A top priority problem of national radiation protection - proper disposal of research reactor spent fuel

International Nuclear Information System (INIS)

Marinkovic, N.; Matausek, M.V.; Jovic, V.

1997-01-01

The paper presents basic facts about RA research reactor at the Vinca Institute. The present state of the RA reactor spent fuel storage pool appears to be a serious safety and radiological problem, which must be solved urgently, independent of the decision about the future status of the reactor itself. The following paragraphs describe current activities on improving storage conditions of the research reactor RA spent fuel. Activities performed so far, concerning identification and improvement of the spent fuel storage conditions are presented. These are verification of radiation protection measures, radiological and chemical analyses, visual inspection and photographing, safety analyses and nuclear criticality studies.A project for long-term solution of the research reactor spent fuel storage is proposed. In order to minimise further corrosion and establish strict control of all the relevant technological parameters of the utility, improvement of conditions for disposal of the fuel in the existing storage, is foreseen in the first phase. New dry storage for long-term storing of the spent fuel should be built during the second phase of the project. Particular attention is paid to the activities related to radiation protection and waste treatment, starting from standard monitoring and control, radiological analyses, regulations and legislation, to complicated handling of high level radioactive waste. (authors)

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

International Nuclear Information System (INIS)

Elnokity, O.E.M.

2006-01-01

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

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

The overpressure protection for the chemical reactors: the batch-size approach

International Nuclear Information System (INIS)

Dellavedova, M.; Gigante, L.; Lunghi, A.; Pasturenzi, C.; Cardillo, P.; Gerosa, N.P.; Rota, R.

2008-01-01

Small and medium enterprises (SMEs) main feature is to run batch and semi-batch processes, working on job orders. They generally have multi propose reactors with an emergency relief system (ERS) already installed. These are normally sized when the reactor is designed, assuming as worst incidental scenario a single phase vapour flow generated by a fire developed outside the apparatus. These assumptions can lead to a big underestimation of the vent area if the actual flow is two-phase and besides generated by a runaway reaction. ERS sizing is particularly hazardous and complex for small mills, as for example fine chemicals and pharmaceutical companies. These factories have usually narrow financial and personal resources, moreover they often use fast processes turnovers. In many cases a complete safety study or the replacement of the ERS is not possible and it can lead to not sustainable costs. The batch-size approach is focused on discontinuous process conditions: aim of this approach is to find the reactor fill level that can lead a vapour single phase flow whether an incident occurs, this condition is considered safe that the ERS installed on the reactor can protect the plant from explosions [it

A coincidencd logic reactor protection system with automatic and permanent testing

International Nuclear Information System (INIS)

Tricornot.

1978-01-01

Within the context of sodium-cooled fast reactors, the CGEE Alsthom enterprise, under consulting for Novatome, has been engaged with the development of an specific protection system for emergency shutdown situations. System described has been conceived on several stages according to the general organization shown on figure I of the Annex, in such a manner that the exigences and recommendations from the safety regulatory authorities are respected and, at the same time, it is assured a significant reactor operation availability without an spureous rod drop. As an example, a selection of principles, rules and criteria currently applied to the development of a system of this kind is reminded. (J.E. de C.)

The Automatic Test Features of the IDiPS Reactor Protection System

International Nuclear Information System (INIS)

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

2007-01-01

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

Twenty years of Radiology in RP-10 nuclear reactor protection; Veinte anos de proteccion radiologica en el reactor nuclear RP-10

Energy Technology Data Exchange (ETDEWEB)

Zapata, Alejandro L.; Ramos, Fernando T.; Arrieta, Rolando W.B.; Vela Mora, Mariano, E-mail: lzapata@ipen.gob.pe, E-mail: framos@ipen.gob.pe, E-mail: rarrieta@ipen.gob.pe, E-mail: mvela@ipen.gob.pe [Instituto Peruano de Energia Nuclear (IPEN), Lima (Peru)

2013-07-01

In this report we present the results about radiation controls during 1990 - 2010, carried out in the Nuclear Reactor RP-10 of the Nuclear Center of Huarangal. These controls and radiological evaluation are of much utility for the radio personnel protection of this one and other reactors, since it allows to compares these variables with respect to the time. From the results obtained in monitoring and radiation controls, we conclude that in no case it has been reached the limits allowed by the Peruvian Regulating Authority. (author)

CHANGE IN DEFORMATION PROPERTIES MODELING OF CONCRETE IN PROTECTIVE STRUCTURES OF NUCLEAR REACTOR BY IONIZING RADIATION

Directory of Open Access Journals (Sweden)

E. K. Agakhanov

2016-01-01

Full Text Available The necessity of studying the effect impact of elementary particles impact on the strength and deformation materials properties used in protective constructions nuclear reactors and reactor technology has been stipulated. A nuclear reactor pressure vessel from prestressed concrete, combining the functions of biological protection is to be considered. The neutron flux problem distribution in the pressure vessel of a nuclear reactor has been solved. The solution is made in axisymmetric with the finite element method using a flat triangular finite element. Computing has been conducted in Matlab package. The comparison with the results has been obtained using the finite difference method, as well as the graphs of changes under the influence of radiation exposure and the elastic modulus of concrete radiation deformations have been constructed. The proposed method allows to simulate changes in the deformation properties of concrete under the influence of neutron irradiation. Results of the study can be used in the calculation of stress-strain state of structures, taking into account indirect heterogeneity caused by the physical fields influence.

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)

Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Louison, R.; Boardman, C.E.

1981-01-01

The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events

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.)

Reliability analysis of protection system of advanced pressurized water reactor - APR 1400

International Nuclear Information System (INIS)

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

2003-04-01

Reliability analysis was carried out for the protection system of the Korean Advanced Pressurized Water Reactor - APR 1400. The main focus of this study was the reliability analysis of digital protection system, however, towards giving an integrated statement of complete protection reliability an attempt has been made to include the shutdown devices and other related aspects based on the information available to date. The sensitivity analysis has been carried out for the critical components / functions in the system. Other aspects like importance analysis and human error reliability for the critical human actions form part of this work. The framework provided by this study and the results obtained shows that this analysis has potential to be utilized as part of risk informed approach for future design / regulatory applications

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

Keats, A.B.

1980-01-01

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

Reactor safety and radiation protection. Draft of the BMU deparmental budget 16 of the 1998 German federal budget

International Nuclear Information System (INIS)

Anon.

1997-01-01

The expenditures earmarked for reactor safety and radiation protection in the 1998 budget of the German Federal Ministry for the Environment, Nature Conservation, and Reactor Safety (BMU) total DM 101 million. The expenditures of the German Federal Office for Radiation Protection (BfS) are to amount to a total of DM 579 million. These are the figures included in departmental budget 16 of the 1998 federal budget, which was discussed by the Federal Parliament in September 1997. The atw compilation singles out a number of significant items of the departmental budget. (orig.) [de

French experience in using programmable systems for the control and the protection of nuclear reactors

International Nuclear Information System (INIS)

Jover, P.

1986-01-01

The paper presents the results obtained in the use of two automated systems important to safety in 1300 MWE French nuclear reactors: the numerical integrated protection system (SPIN) and the logical control system (CONTROBLOC)

International guidelines for fire protection at nuclear installations including nuclear fuel plants, nuclear fuel stores, teaching reactors, research establishments

International Nuclear Information System (INIS)

The guidelines are recommended to designers, constructors, operators and insurers of nuclear fuel plants and other facilities using significant quantities of radioactive materials including research and teaching reactor installations where the reactors generally operate at less than approximately 10 MW(th). Recommendations for elementary precautions against fire risk at nuclear installations are followed by appendices on more specific topics. These cover: fire protection management and organization; precautions against loss during construction alterations and maintenance; basic fire protection for nuclear fuel plants; storage and nuclear fuel; and basic fire protection for research and training establishments. There are numerous illustrations of facilities referred to in the text. (U.K.)

Enhancement the physical protection system of the WWR-SM reactor at Institute of Nuclear Physics of Academy of Science of the Republic of Uzbekistan

International Nuclear Information System (INIS)

Karabaev, Kh.Kh.; Rakhimbaev, A.T.; Rakhmanov, A.B.; Salikhbaev, U.S.; Yuldashev, B.S.

2004-01-01

Full text: Joining of the Republic of Uzbekistan to Non-Proliferation Treaty required the revision of nuclear fuel protection system and radioactive sources from illegal access in all stages of work with nuclear materials. One of the primary technical actions of ensuring non-proliferation of nuclear materials is physical protection. The project was worked out on upgrading and enhancement of the physical protection of the reactor building. In cooperation with Sandia National Laboratory and support of the Department of Energy (DOE) USA The first stage of the physical protection upgrading provided for fresh fuel protection: - the new fresh fuel storage room was built and equipped with the modern control and detection system, - the reactor building was equipped with detection devices and access control, - the central alarm station (CAS) has been built and equipped with computer control and observing system, - code access system has been implemented. The first stage of upgrading of physical protection system was accomplished for 4 months, and put into operation in 1996. The second stage of physical protection system modernization included the construction of the second barrier of the physical protection, equipping it with observation and control devices and also extension of the CAS. The perimeter around the reactor building was cleaned from trees, bushed and in a short time a two-fence barrier was erected. The access control point provided the secured intensified control of the access to the reactor territory. The physical protection system was supplied with equipment for safeguard and TV observation of perimeter, access control to the territory of the reactor: - the CAS was extended and computer observation control system was upgraded, - the badge station has been constructed, equipped and set up, - all doors, windows, reactor hall gate have been replaced by strengthened metal ones, - uninterruptible power supply (UPS) and diesel-generator have been installed, - the

Study on human-factors-engineering properties of reactor maintenance workers with protection suits, (2). Basic research on various biological characteristics in reactor maintenance simulation tests

Energy Technology Data Exchange (ETDEWEB)

Yoshino, K; Ishii, K; Nakasa, H [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Shigeta, S

1980-11-01

To ensure the safety of reactor maintenance workers and to reduce the radiation exposure through the enhancement of labor efficiency, it is needed to evaluate quantitatively work-stress levels of workers with radiation-protection suits. This paper presents the results of reactor-maintenance simulation tests in which the relationship between the work stress and biological characteristics is investigated for 5 pinds of model works done by testees without protection suits in an artificial climate chamber. Major results obtained are: (1) the selected model works are mostly evaluated to be relatively heavy through the measurement of RMR (Relative Metabolic Rate). (2) biological characteristics such as heart rate and respiratory volume under the model works have close relationship to RMR which is the cumulative quantity in relatively long time, and then they may become the real-time indicator for the work stress level. (3) such biological characteristics are greatly affected by the high-temperature work-environment which is often seen in workers with protection suits.

Development of the signalling circuits for reactor emergency protection systems

International Nuclear Information System (INIS)

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

1978-01-01

Construction of circuits for nuclear reactor emergency protection according to the power level and rate of power rise with the use of integrated microcircuits is discussed. Circuits of relay- and transformer-based logical signaling devices are presented. It is noted that disadvantages of a transformer-based loaical sianaling device are great power consumption (about 300 mW) and slow response limited by the time constant of the output smoothing filter. Further development of circuits under consideration is associated with the employment of new optronic elements intended to replace the transformers

Possibilities of achieving non-positive void reactivity effect in fast sodium-cooled reactors with increased self-protection

International Nuclear Information System (INIS)

Alekseev, P.N.; Zverkov, Yu.A.; Morozov, A.G.; Orlov, V.V.; Slesarev, I.S.; Subbotin, S.A.

1989-01-01

The problems of self-protection inhancement for the liquid-metal cooled fast reactors with intra-assembly heterogeneity of the core are studied. Possible approaches to arrangement of such reactors with various powers characterized by high levels of coolant natural circulation, minimum reactivity changes during fuel burn-up and non-positive void effect of reactivity are found. 10 refs.; 11 figs

United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

International Nuclear Information System (INIS)

1997-06-01

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

International Nuclear Information System (INIS)

Parkes, Alison; Saratsopoulos, John; Deura, Michael; Kenny, Pat

2008-01-01

The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

Method for performing diversity and defense-in-depth analyses of reactor protection systems

International Nuclear Information System (INIS)

Preckshot, G.G.

1994-12-01

The purpose of this NUREG is to describe a method for analyzing computer-based nuclear reactor protection systems that discovers design vulnerabilities to common-mode failure. The potential for common-mode failure has become an important issue as the software content of protection systems has increased. This potential was not present in earlier analog protection systems because it could usually be assumed that common-mode failure, if it did occur, was due to slow processes such as corrosion or premature wear-out. This assumption is no longer true for systems containing software. It is the purpose of the analysis method described here to determine points of a design for which credible common-mode failures are uncompensated either by diversity or defense-in-depth

Electromagnetic drive of the control and protection system of a nuclear reactor

International Nuclear Information System (INIS)

Zav'yalova, G.I.

1983-01-01

The design and operating principle of an electromagnetic drive with a linear synchronous reaction motor are described. At the present time, electromagnetic control mechanisms using linear electric motors are finding increasingly widespread application as drives for the control and protection system of nuclear reactors. In these drives there is a functional mergence of the electromagnetic mechanism with the final control element; these drives, therefore, have advantages over electromechanical drives

Using a research reactor to teach practical radiation protection

International Nuclear Information System (INIS)

Musilek, A.; Steinhauser, G.

2010-01-01

To teach students about the practical handling of radioactive materials and the related radiation protection, it is advantageous to be able to produce radioactive material with specific properties. Through the neutron activation of specific samples, radio-nuclides can be produced that are precisely tailored for particular experiments, both in type of radiation (beta, gamma) as well as in activity and half-life. At the Atominstitut in Vienna, a 250 kW TRIGA Mark II research reactor is used for the production of these nuclides. In this paper, four practical exercises are presented, covering many questions and challenges that occur in radiation protection. The first exercise uses neutron activation of sodium-chloride to cover theoretical aspects of the calculation of dose rates (using dose rate constants) through the activation of Na-23, Cl-35 and Cl-37 (including cross sections, half-life, inverse square law), as well as a practical examination (handling of dose rate meters). The second exercise gives students the opportunity to decontaminate a laboratory after an incident under realistic circumstances. For this exercise, KNO 3 is activated in the reactor. The resulting K-nuclide produces no risk of inadequate decontamination for the laboratory, since the half-life of K-42 is only 12 h. The third exercise is designed to teach students how to deal with unsealed radioactive material by irradiation of ammonium dihydrogenphosphate. In this case, an only-beta-active (P-32) fertilizer is produced, which is applied to plants in subsequent chemical processing. In the following step, the 'quality of this fertilizer' is determined by measuring the absorbed activity of the plant leaves using a GM counter. The fourth exercise is another approach in working with unsealed radioactive material. It simulates the PUREX process to separate uranium from fission products using a liquid - liquid extraction. (authors)

Protective interior wall and attaching means for a fusion reactor vacuum vessel

International Nuclear Information System (INIS)

Phelps, R.D.; Upham, G.A.; Anderson, P.M.

1988-01-01

A protective wall for the interior surface of a fusion reactor vessel wall is described comprising: an array of plates, each plate of the array including a main body section, a pair of edge sections bent at an angle with respect to the main body section, and a pair of flange-like end sections each having protruding sections with cut-aways therein, the protruding sections of the flange-like end sections extending in a direction substantially parallel to the main body section; and means operatively associated with the protruding sections of the flange-like end sections of the plates for mounting the array of plates to an associated vessel wall to be protected

Extensions and renovations of reactor protection systems

International Nuclear Information System (INIS)

Hellmerichs, K.

1985-01-01

Increase of requirements by the authorities as to the design of reactor protection systems affected in the last years not only plans being under construction, but also resulted in partly spacious extensions and renovations. While working on the extensions and renovations a lot of problems arose: far-reaching performance of newest guidelines and rules in spite of old plant concepts; partly higher degree of redundancy requirements of the new systems in contrast to the present systems; use of present safeguard systems for new accident countermeasures; designation of priorities between present and new functions, especially in view of fault behaviour of present systems; adaptation of the new I and C equipment to the present signalisation-, operation- and information-arrangements under consideration of the present operational philosophy; spatial incorporation of new equipments; construction as to time without expanding of the planned refuelling phases. Because the KWU has planned and constructed such alterations in nearly 10 plants a lot of experience has been gathered. (author)

Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

Directory of Open Access Journals (Sweden)

Hyun-Kook Shin

2006-08-01

Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

Nuclear reactor safety protection device

International Nuclear Information System (INIS)

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

1994-01-01

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

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Safety-technical comparison of a hard-wired dynamic reactor protection system with a computerized alternative. Pt. 2

International Nuclear Information System (INIS)

Buettner, W.E.

1978-03-01

The investigation presented here compares a conventional hard-wired dynamic reactor protection system with a computerized alternative. For both systems the spurious trip probability is determined, i.e. the probability of a false release of one or more protection actions due to failures. The mean unavailability is also determined for those not distinctly safety-oriented protection actions of the computerized protection system which are released via the open circuit current principle. This is because system breakdowns prevent the release of those protection actions in case of demand. The advantages and disadvantages of either type of system are viewed against each other. (orig.) [de

Device for protecting the containment vessel dome of a nuclear reactor

International Nuclear Information System (INIS)

Allain, A.; Filloleau, E.; Mulot, P.

1976-01-01

A device is disclosed for protecting the dome of a nuclear reactor containment vessel against the upward displacement of the concrete shield slab of said reactor and the resultant effects of tilting of an equipment unit mounted on the shield slab at the periphery of said slab, wherein said device comprises: (1) means for separating the equipment unit into two sections consisting of an upper section and a lower section, said lower section being rigidly fixed to said shield slab and said means being actuated by the upward displacement of said slab, (2) a system for vertical rectilinear guiding of said upper section within the containment vessel, and (3) rigid mechanical components which provide a coupling between the aforesaid upper and lower sections of the equipment unit and exert on said upper section under the action of the tilting motion of said lower section a thrust which causes the upward displacement of said upper section

Radiation protection at the RA Reactor in 1993, Part II, Decontamination and actions, collection of liquid effluents and solid radioactive waste

International Nuclear Information System (INIS)

Mandic, M.; Vukovic, Z.; Lazic, S.; Plecas, I.; Voko, A.

1993-01-01

Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

Development of an Advanced Digital Reactor Protection System Using Diverse Dual Processors to Prevent Common-Mode Failure

International Nuclear Information System (INIS)

Shin, Hyun Kook; Nam, Sang Ku; Sohn, Se Do; Chang, Hoon Seon

2003-01-01

The advanced digital reactor protection system (ADRPS) with diverse dual processors has been developed to prevent common-mode failure (CMF). The principle of diversity is applied to both hardware design and software design. For hardware diversity, two different types of CPUs are used for the bistable processor and local coincidence logic (LCL) processor. The Versa Module Eurocard-based single board computers are used for the CPU hardware platforms. The QNX operating system and the VxWorks operating system were selected for software diversity. Functional diversity is also applied to the input and output modules, and to the algorithm in the bistable processors and LCL processors. The characteristics of the newly developed digital protection system are described together with the preventive capability against CMF. Also, system reliability analysis is discussed. The evaluation results show that the ADRPS has a good preventive capability against the CMF and is a highly reliable reactor protection system

Control of WWER-440 nuclear reactor

International Nuclear Information System (INIS)

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

1978-01-01

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

Reliability analysis for Atucha II reactor protection system signals

International Nuclear Information System (INIS)

Roca, Jose Luis

1996-01-01

Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

Reliability analysis for Atucha II reactor protection system signals

International Nuclear Information System (INIS)

Roca, Jose L.

2000-01-01

Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

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

On implementation of the self-protection principle to the reactors with fast-resonance neutron spectrum

International Nuclear Information System (INIS)

Kuznetsov, V.V.; Morozov, A.G.; Slesarev, I.S.; Alekseev, P.N.; Zverkov, Yu.A.; Subbotin, S.A.

1990-01-01

The calculational substantiation of SWPR posessing inherent physical properties of self-protection against possible accidents is given. A variety of approaches to the layout of these reactors have been found, the possible level of their fuel utilization characteristics is analyzed. 12 refs.; 14 figs.; 6 tabs

Solution to the incompatibility between reactor protection logic and turbine shot logic. Scram by high pressure; Solucion a la incompatibilidad entre logica de proteccion de reactor y logica de disparo de turbina. SCRAM por alta presion

Energy Technology Data Exchange (ETDEWEB)

Ramos Q, R.; Santiago F, C.; Gonzalez P, G., E-mail: ruben.ramos01@cfe.gob.mx [Comision Federal de Electricidad, Central Nuclear Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

2013-10-15

The nuclear power plant of Laguna Verde carried out the Modernization and Increase of Extended Power Project in its two Units (2005-2011). This modernization included to the electro-hydraulic control system of the main turbine, replacing an ana logical system by one digital (Digital Electro-hydraulic Control - DEHC) whose functions are of controlling the reactor pressure in the different operation ways as wells as of controlling the velocity and load of the main turbine. Also, it has protections that are related with diverse plant systems, as the Reactor Protection Systems (RPS). During the tests stage was realized a programmed load rejection, which Reactor Scram should cause when being presented the shot of main turbine. However, the logic of the RPS was inhibited due to the quick response of the new control DEHC, propitiating a condition of non prospective plant and, in consequence, the Reactor Scram happened for another protection of the RPS. (author)

Safety assessment principles for reactor protection systems in the United Kingdom

International Nuclear Information System (INIS)

Philp, W.

1990-01-01

The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems

Safety assessment principles for reactor protection systems in the United Kingdom

Energy Technology Data Exchange (ETDEWEB)

Philp, W

1990-07-01

The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems.

Optimum supervision intervals and order of supervision in nuclear reactor protective systems

International Nuclear Information System (INIS)

Kontoleon, J.M.

1978-01-01

The optimum inspection strategy of an m-out-of-n:G nuclear reactor protective system with nonidentical units is analyzed. A 2-out-of-4:G system is used to formulate a multi-variable optimization problem to determine (a) the optimum order of supervision of the units and (b) the optimum supervision intervals between units. The case of systems with identical units is a special case of the above. Numerical results are derived using a computer algorithm

Environment report 1990 of the Federal Minister for the Environment, Nature Protection and Reactor Safety

International Nuclear Information System (INIS)

1990-01-01

The 'Environment Report 1990' describes the environmental situation in the Federal Republic of Germany; draws a balance of environmental policy measures taken and introduced; gives information on future fields of action in environmental policy. The 'Environment Report 1990' also deals with the 'Environment Expert Opinion 1987', produced by the board of experts on environmental questions. It contains surveys of the following sectors: Protection against hazardous materials air pollution abatement, water management, waste management, nature protection and preservation of the countryside, soil conservation, noise abatement, radiation protection, reactor safety. A separate part of the 'Environment Report 1990' deals with the progress made in 'interdisciplinary fields' (general law on the protection of the environment, instruments of environmental policy, environmental information and environmental research, transfrontier environmental policy). (orig./HP) [de

Preliminary Validation and Verification Plan for CAREM Reactor Protection System

International Nuclear Information System (INIS)

Fittipaldi, Ana; Maciel Felix

2000-01-01

The purpose of this paper, is to present a preliminary validation and verification plan for a particular architecture proposed for the CAREM reactor protection system with software modules (computer based system).These software modules can be either own design systems or systems based in commercial modules such as programmable logic controllers (PLC) redundant of last generation.During this study, it was seen that this plan can also be used as a validation and verification plan of commercial products (COTS, commercial off the shelf) and/or smart transmitters.The software life cycle proposed and its features are presented, and also the advantages of the preliminary validation and verification plan

Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10

Directory of Open Access Journals (Sweden)

Chengxiang Guo

2016-01-01

Full Text Available Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR. To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10 in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% for γ dose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

Depth protection system

International Nuclear Information System (INIS)

Arita, Setsuo; Izumi, Shigeru; Suzuki, Satoru; Noguchi, Atomi.

1988-01-01

Purpose: To previously set a nuclear reactor toward safety side by the reactor scram if an emergency core cooling system is failed to operate. Constitution If abnormality occurs in an emergency core cooling system or an aqueous boric acid injection system, a reactor protection system is operated and, if the reactor protection system shows an abnormal state, a control rod withdrawal inhibition system is operated as a fundamental way. For instance, when the driving power source voltage for the emergency core cooling system is detected and, if it is lower than a predetermined value, the reactor protection system is operated. Alternatively, if the voltage goes lower than the predetermined value, the control rod withdrawal is inhibited. In addition, stopping for the feedwater system is inhibited. Further, integrity of the driving means for the emergency core cooling system is positively checked and the protection function is operated depending on the result of check. Since the nuclear reactor can be set toward the safety side even if the voltage for the driving power source of the aqueous boric acid injection system is lower than a predetermined value, the reactor safety can further be improved. (Horiuchi, T.)

RA Research nuclear reactor, Part II - radiation protection at the RA nuclear reactor in 1984; Istrazivacki nuklearni reaktor RA - Deo II - zastita od zracenja kod nuklearnog reaktora RA u 1984. godini

Energy Technology Data Exchange (ETDEWEB)

Ninkovic, M; Ajdacic, N; Zaric, M; Vukovic, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1984-12-15

Radon protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each category is described as a separate annex of this report.

Utilization of a statistical procedure for DNBR calculation and in the survey of reactor protection limits

International Nuclear Information System (INIS)

Pontedeiro, A.C.; Camargo, C.T.M.; Galetti, M.R. da Silva.

1987-01-01

A new procedure is applied to Angra 1 NPP, which is related to DNBR calculations, considering the design parameters statistically: Improved Thermal Design Procedure (ITDP). The ITDP application leads to the determination of uncertainties in the input parameters, the sensitivity factors on DNBR. The DNBR limit and new reactor protection limits. This was done to Angra 1 with the subchannel code COBRA-IIIP. The analysis of limiting accident in terms of DNB confirmed a gain in DNBR margin, and greater operation flexibility of the plant, decreasing unnecessary trips of the reactor. (author) [pt

Modelling of reactor control and protection systems in the core simulator program GARLIC

International Nuclear Information System (INIS)

Beraha, D.; Lupas, O.; Ploegert, K.

1984-01-01

For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

Radiation protection at the RA Reactor in 1989, Part -2, Decontamination, collection of treatment of fluid and solid radioactive waste, Annex 3

International Nuclear Information System (INIS)

Mandic, M.; Vukovic, Z.; Plecas, I.; Knezevic, Lj.; Lazic, S.; Bacic, S.

1989-01-01

Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

RA Research reactor, Annual report 1988

International Nuclear Information System (INIS)

Sotic, O.

1988-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Research and engineering application of coordinated instrumentation control and protection technology between reactor and steam turbine generator on nuclear power plant

International Nuclear Information System (INIS)

Sun Xingdong

2014-01-01

The coordinated instrumentation control and protection technology between reactor and steam turbine generator (TG) usually is very significant and complicated for a new construction of nuclear power plant, because it carries the safety, economy and availability of nuclear power plant. Based on successful practice of a nuclear power plant, the experience on interface design and hardware architecture of coordinated instrumentation control and protection technology between reactor and steam turbine generator was abstracted and researched. In this paper, the key points and engineering experience were introduced to give the helpful instructions for the new project. (author)

Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

1981-01-01

Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

1981-01-01

Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

1981-01-01

Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Physical protection of shipments of irradiated reactor fuel; Interim guidance. Regulatory report

International Nuclear Information System (INIS)

1980-06-01

During May, 1979, the U.S. Nuclear Regulatory Commission approved for issuance in effective form new interim regulations for strengthening the protection of spent fuel shipments against sabotage and diversion. The new regulations were issued without benefit of public comment, but comments from the public were solicited after the effective date. Based upon the public comments received, the interim regulations were amended and reissued in effective form as a final interim rule in May, 1980. The present document supersedes a previously issued interim guidance document, NUREG-0561 (June, 1979) which accompanied the original rule. This report has been revised to conform to the new interim regulations on the physical protection of shipments of irradiated reactor fuel which are likely to remain in effect until the completion of an ongoing research program concerning the response of spent fuel to certain forms of sabotage, at which time the regulations may be rescinded, modified or made permanent, as appropriate. This report discusses the amended regulations and provides a basis on which licensees can develop an acceptable interim program for the protection of spent fuel shipments

Radiation protection at the RA Reactor in 1995, Part -2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste

International Nuclear Information System (INIS)

Mandic, M.; Vukovic, Z.; Lazic, S.; Plecas, I.; Voko, A.

1995-01-01

Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

Radiation protection at the RA Reactor in 1998, Part 2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste

International Nuclear Information System (INIS)

Mandic, M.; Vukovic, Z.; Bacic, S.; Plecas, I.

1998-01-01

Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

Optimisation of radiation protection for the new european pressurized water reactor (EPR)

International Nuclear Information System (INIS)

Miniere, D.; Beneteau, Y.; Le Guen, Bernard

2008-01-01

Full text: As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection (RP) aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing RP-significant activities has been to include all involved stake holders -designers, licensee and contractor companies- in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The RP target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 man.Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the EPR 's reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), RP studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the accessible area can be posted as a green area (dose rate < 25

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Diversity in computerized reactor protection systems

International Nuclear Information System (INIS)

Fischer, H.D.; Piel, L.

1999-01-01

Based on engineering judgement, the most important measures to increase the independency of redundant trains of a computerized safety instrumentation and control system (I and C) in a nuclear power plant are evaluated with respect to practical applications. This paper will contribute to an objective discussion on the necessary and justifiable arrangement of diversity in a computerized safety I and C system. Important conclusions are: - (i) diverse equipment may be used to control dependent failures only if measures necessary for designing, licensing, and operating a computerized safety I and C system homogeneous in equipment are neither technically nor economically feasible; - (ii) the considerable large operating experience in France with a non-diverse equipment digital reactor protection system does not call for equipment diversity. Although there are no generally accepted methods, the licensing authority is still required to take into account dependent failures in a probabilistic safety analysis; - (ii) the frequency of postulated initiating events implies which I and C functionality should be implemented on diverse equipment. Using non-safety I and C equipment in addition to safety I and C equipment is attractive because its necessary unavailability to control an initiating event in teamwork with the safety I and C equipment is estimated to range from 0.01 to 0.1. This can be achieved by operational experience

Development of digital plant protection system for Korean Next Generation Reactor

International Nuclear Information System (INIS)

Suk-Joon Park

1998-01-01

A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author)

Cyber Security Analysis by Attack Trees for a Reactor Protection System

International Nuclear Information System (INIS)

Park, Gee-Yong; Lee, Cheol Kwon; Choi, Jong Gyun; Kim, Dong Hoon; Lee, Young Jun; Kwon, Kee-Choon

2008-01-01

As nuclear facilities are introducing digital systems, the cyber security becomes an emerging topic to be analyzed and resolved. The domestic and other nation's regulatory bodies notice this topic and are preparing an appropriate guidance. The nuclear industry where new construction or upgrade of I and C systems is planned is analyzing and establishing a cyber security. A risk-based analysis for the cyber security has been performed in the KNICS (Korea Nuclear I and C Systems) project where the cyber security analysis has been applied to a reactor protection system (RPS). In this paper, the cyber security analysis based on the attack trees is proposed for the KNICS RPS

RA Research nuclear reactor - Annual report 1987

International Nuclear Information System (INIS)

1987-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Research nuclear reactor RA - Annual Report 1989

International Nuclear Information System (INIS)

Sotic, O.

1989-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

IDRA 2016 Technical FAQs Word doc_FR

International Development Research Centre (IDRC) Digital Library (Canada)

Centre Awards

Cela prend un certain temps pour remplir le formulaire de demande en ligne. Le système vous permet de travailler à votre propre rythme. Vous pouvez téléverser un document, le sauvegarder, puis continuer plus tard. Le formulaire de demande comporte plusieurs pages, et il est possible de le remplir graduellement.

Radiation protection for repairs of reactor's internals at the 2nd Unit of the Nuclear Power Plant Temelin

International Nuclear Information System (INIS)

Zapletal, P.; Konop, R.; Koc, J.; Kvasnicka, O.; Hort, M.

2011-01-01

This presentation describes the process and extent of repairs of the 2 nd unit of the Nuclear power plant Temelin during the shutdown of the reactor. All works were optimized in terms of radiation protection of workers.

A bustling academic reactor creates challenges and opportunities in the area of physical protection

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.; Standley, V.

2001-01-01

The 250 kW TRIGA research reactor is located at the Atominstitut Vienna, Austria, only a few subway stations from the city centre of Vienna. Its main purpose is the training of university students in the field of nuclear engineering and radiation protection as well as in radiochemistry and neutron- and solid-state physics. The existing facility is visited during a normal academic year by about 300 persons per day falling into seven different categories including fully employed staff, students occasionally visiting a seminar, and IAEA personnel from all over the world. These different groups have to be accounted for daily and are separated into different categories in view of security and physical protection. (author)

Radiation protection at the RA reactor in 1984, Part II c Radioactivity control in the environment of the RA reactor, Meteorology measurements

International Nuclear Information System (INIS)

Grsic, Z.; Zaric, M.; Nikolic, R.; Stevanovic, M.

1984-01-01

During 1984, meteorology measurements were continued as a part of the environmental control of the Vinca Institute. This report covers the period from December 1983 - November 1984. Part of the meteorology measurements and data analysis is adapted to the needs of the Institute, i.e. RA reactor and some Laboratories. The objective of these activities is forming the data base for solving everyday and special problems related to control, protection and safety of Institute environment

The implementation and evaluation of physical protection system of the IEA-R1 reactor

International Nuclear Information System (INIS)

Vaz, Antonio Carlos Alves

2016-01-01

The September 11, 2001 terrorist attacks in New York, the accident at the Fukushima nuclear power plant on March 2011 and the recent attacks in Paris on November 2015 are examples of events that justify the efforts of the International Agency of Energy Atomic - IAEA to improve security at nuclear facility. The Brazilian government has been collaborating with this project and investing resources to improve the Physical Protection System - PPS of the nuclear research reactor system, technically is associated with the elements of detection, delay and response. The PPS is an integrated system of people, equipment and procedures used to protect nuclear facilities and radioactive sources against threat, theft or sabotage. The PPS works to avoid, to mitigate or to minimize the consequences caused by these actions. This study evaluates the PPS of the reactor, identifying the vulnerabilities and suggesting ways to improve the system effectiveness. The analyses were based on the methodology developed by Sandia National Laboratories´ security experts in Albuquerque - USA, allowing the system evaluation through hypothetical and probabilistic analyzes; identifying threats, determining the targets and analyzing the possible adversaries paths. From the methodology adopted was obtained the value around 40% for PE indicator, which shows the need to improve the system to minimizing the vulnerabilities. (author)

Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

International Nuclear Information System (INIS)

1986-11-01

Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

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

Off-site protective action selection for nuclear reactor accidents

International Nuclear Information System (INIS)

Weerakkody, S.D.

1986-01-01

A computer program based upon a model using a rational theoretical basis was developed to select appropriate off-site protective actions during nuclear reactor accidents. The special features of this program include (a) introduction of a precursor concept that uses the history of the accident progression to determine the spectrum of potential accident scenarios and estimates of the likelihoods of each accident scenario; (b) use of statistical decision theory and the concept of entropy of a spectrum to select the appropriate protective actions using either the minimax principle or the Bayes action method; and (c) introduction of methods to quantify evacuation travel risks. In order to illustrate the usefulness of the computer program, it was applied at three stages of the Three Mile Island accident scenario. Quantified non-radiological risks of evaluation have been used to establish dose thresholds below which evacuations are not justified. Using the Poisson analysis for evacuation risks and the absolute L-L BEIR model for radiation risk suggests 330 mrems as a reasonable value for this threshold. The usefulness of the program in developing a technical basis to select the size of the plume exposure pathway emergency planning zone (EPZ) is discussed. Quantified evacuation risks, cost, and the current rationale upon which the EPZ is based, justify an EPZ between 5-10 miles for WASH-1400 source-terms

Development of digital plant protection system for Korean Next Generation Reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Suk-Joon [NSSS Engineering and Development Division, Korea Power Engineering Company, Taejon (Korea, Republic of)

1998-10-01

A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author) 8 refs, 4 figs, 3 tabs

Cyber Security Analysis by Attack Trees for a Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Park, Gee-Yong; Lee, Cheol Kwon; Choi, Jong Gyun; Kim, Dong Hoon; Lee, Young Jun; Kwon, Kee-Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

As nuclear facilities are introducing digital systems, the cyber security becomes an emerging topic to be analyzed and resolved. The domestic and other nation's regulatory bodies notice this topic and are preparing an appropriate guidance. The nuclear industry where new construction or upgrade of I and C systems is planned is analyzing and establishing a cyber security. A risk-based analysis for the cyber security has been performed in the KNICS (Korea Nuclear I and C Systems) project where the cyber security analysis has been applied to a reactor protection system (RPS). In this paper, the cyber security analysis based on the attack trees is proposed for the KNICS RPS.

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Pilot program: NRC severe reactor accident incident response training manual: Public protective actions: Predetermined criteria and initial actions

International Nuclear Information System (INIS)

Martin, J.A. Jr.; McKenna, T.J.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Public Protective Actions - Predetermined Criteria and Initial Actions is the fourth in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume reviews public protective action criteria and objectives, their bases and implementation, and the expected public response. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

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

Nuclear Reactor RA Safety Report, Vol. 9, Radiation protection; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 9, Zastita od zracenja

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-11-01

Instrumentation for Radiation protection existing at the RA reactor is dating mostly from the period 1957-1959 when the reactor has been built. With some minor exception it was produced in USSR. Radiation protection system was constructed based on specific design project, somewhat modified original USSR project which has been indispensable because of some modification of the building design. During the past 27 years no renewal of the instrumentation was done, only maintenance was performed. Instrumentation consists of old electronic devices which caused difficulties and even prevented regular maintenance because of lack of spare parts. Instrumentation for radiation protection at the RA reactor is classified as follows: centralized dosimetry system; stationary dosimetry instrumentation, movable and personal dosimetry systems. Apart from the scheme of dosimetry instrumentation this volume includes description of radiation protection procedures; protection devices; radiation doses and dose limit data; program for environmental radioactivity control; medical control procedures. [Serbo-Croat] Instrumentacija za zastitu od zracenja koja danas postoji na reaktoru RA najvecim delom potice iz perioda 1957-1959 kada je reaktor gradjen. Sa malim izuzetcima instrumentacija je proizvedena u SSSR. Sistem za zastitu izveden je na osnovu posebnog projekta koji predstavlja modifikaciju originalnog projekta koja je bila neophodna usled modifikovanog gradjevinskog projekta. U proteklom periodu od 27 godina instrumentacija nije obnavljana vec je vrseno samo odrzavanje. Instrumentacija je izradjena u danas prevazidjenoh tehnoligiji (elektronske cevi) sto je otezavalo i skoro onemogucavalo normalno odrzavanje. Instrumentacija za zastitu od zracenja na reaktoru RA moze se podeliti na tri dela: centralizovani dozimetrijski sistem; stacionarna dozimetrijska instrumentacija; prenosna i licna dozimetrijska instrumentacija. Pored seme sistema dozimetrijske kontrole ova knjiga sadrzi opis

Structure of thermonuclear reactor wall

International Nuclear Information System (INIS)

Yamazaki, Seiichiro.

1991-01-01

In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

Dynamic response of aircraft impact of a reactor building with protective shell on independent foundation

International Nuclear Information System (INIS)

Constantopoulos, I.V.; Vardanega, C.; Attalla, I.

1981-01-01

Aircraft impact loading can penalize significantly the design of the equipment in a conventional containment building. An alternative scheme was developed in an attempt to reduce the aircraft impact response. A preliminary study was carried out to investigate the feasibility of the alternative scheme. This study was made in such perspective and for the purpose of comparing the response to aircraft impact of a standard reactor building, to that of a reactor building having an independently founded outer shell. In the second scheme, the outer shell is meant to receive the aircraft impact, so that the load will be transmitted to the reactor building internals only by way of the structure-soil-structure system. In both cases, the aircraft impact was postulated to occur on a linear single degree of freedom oscillator which modeled, approximately, the plastification of the impact area. The soil was considered as a half-space with properties corresponding to a medium stiff soil, and modeled by lumped soil springs and dashpots. The reactor internals, inner shell and protective outer shell were modeled with beam elements and concentrated inertias. In modeling the coupled system, soil-structure interaction and structure-to-structure interaction through the soil were represented by a global stiffness matrix corresponding to the three degrees the freedom of each foundation, i.e. horizontal, vertical and rocking. (orig./HP)

The protection system of nuclear reactors. Example of a very high reliability system. Its evolution

International Nuclear Information System (INIS)

Weill, J.

1980-06-01

The present state of reactor protection is described and mention is made of certain evolutionary trends towards completely automated systems which either help the operator to take decisions in the event of an accident or take and execute these decisions for him. To do so, the use of models and recourse to complex data processing systems is necessary. This ensemble reflects an evolution of the reliability of the equipment towards that of the software [fr

High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

International Nuclear Information System (INIS)

1983-01-01

The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

Reliability analysis of reactor protection systems

International Nuclear Information System (INIS)

Alsan, S.

1976-07-01

A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)

RB research nuclear reactor, Annual report for 1981

International Nuclear Information System (INIS)

Markovic, H.; Sotic, O.; Pesic, M.; Vranic, S.; Zivkovic, B.; Bogdanovic, M.; Petronijevic, M.

1981-01-01

The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report

Improved core protection calculator system algorithm

International Nuclear Information System (INIS)

Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

2009-01-01

Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

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)

The Steam Generating Heavy Water Reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

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

Refurbishment of the reactor protection system at Paks NPP. The refurbishment process

International Nuclear Information System (INIS)

Turi, T.; Katics, B.

1998-01-01

The Reactor Protection System Refurbishment Project had an extensive preparation period in Paks started in 1992. During this preparation a large volume of the basic engineering tasks has been performed and as a result a contract for implementation of a three-train digital RPS on the four Units was concluded with Siemens in September, 1996. According to that contract the first refurbished Unit will be commissioned in 1999 followed by a further Unit in each succeeding year. This paper introduces the process of the refurbishment, overview of the V and V activities, introduce the architecture, summarize the main design principles and outlines the additional tasks to be performed together with the RPS design. (author)

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

Protected plutonium breeding by transmutation of minor actinides in fast breeder reactor

International Nuclear Information System (INIS)

Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi

2008-01-01

The improvement of proliferation resistance properties of Pu and the burnup characteristics of fast breeder reactor (FBR) had been studied by utilizing minor actinides (MAs) to produce more 238 Pu from 237 Np and 241 Am through neutron capture reaction. The higher the 238 Pu content in the fuel, the higher the proliferation resistance of the fuel would be owing to the natural characteristics of 238 Pu with high decay heat and high neutron production. The present paper deals with the assessment of passive measure against nuclear material proliferation by focusing on improving the inherent proliferation barrier of discharged Pu from an FBR. Results showed that 5% MA doping to the blanket of an FBR gives as high as 17-19% 238 Pu, which could be seen as a significant improvement of the proliferation properties of Pu. Moreover, additional 5% ZrH 2 , together with 5% MA doping to the blanket, could enhance the 238 Pu fraction much more (22-24%). With an assumption of protected Pu whose 238 Pu isotopic fraction is more than 12%, the present paper revealed that protected Pu could be produced more than the Pu consumed (protected Pu breeding) through incineration in an FBR with doping of a minimum 3% MAs or (2% MAs+5% ZrH 2 ) to the blanket. (author)

Nuclear reactors for the future

International Nuclear Information System (INIS)

Vijayan, P.K.; Kamble, M.T.; Dulera, I.V.

2013-01-01

For the sustainable development of nuclear power plants with enhanced safety features, economic competitiveness, proliferation resistance and physical protection, several advanced reactor developments have been initiated world-wide. The major advanced reactor initiatives and the proposed advanced reactor concepts have been briefly reviewed along with their advantages and challenges. Various advanced reactor designs being pursued in India have also been briefly described in the paper. (author)

Mines vs reactors: comparison of radiation problems and solutions

International Nuclear Information System (INIS)

Bush, W.R.

1981-01-01

Radiation protection in uranium mines and nuclear reactors is compared, with the objective of determining if any radiation protection practices which have worked well in one area can be adapted to the other area, especially from reactors to mines since radiation protection is more highly developed in reactors. Several areas are identified where improvements can be made in radiation protection in mines, the most important being improvement in ventilation at the workplace, greater attention to special monitoring during upset conditions, and a need for a substantial increase in radiation protection staff. It is concluded that, although considerable room exists for improvement, the general approach to radiation protection presently being followed seems appropriate for the circumstances encountered in mining

Regulations for RA reactor operation

International Nuclear Information System (INIS)

1980-09-01

Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

Nuclear reactor power supply

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

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

Radiological protection issues arising during and after the Fukushima nuclear reactor accident

International Nuclear Information System (INIS)

González, Abel J; Akashi, Makoto; Sakai, Kazuo; Yonekura, Yoshiharu; Boice Jr, John D; Chino, Masamichi; Homma, Toshimitsu; Ishigure, Nobuhito; Kai, Michiaki; Kusumi, Shizuyo; Lee, Jai-Ki; Menzel, Hans-Georg; Niwa, Ohtsura; Yamashita, Shunichi; Weiss, Wolfgang

2013-01-01

Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with ‘contamination’ of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of

Radiological protection issues arising during and after the Fukushima nuclear reactor accident.

Science.gov (United States)

González, Abel J; Akashi, Makoto; Boice, John D; Chino, Masamichi; Homma, Toshimitsu; Ishigure, Nobuhito; Kai, Michiaki; Kusumi, Shizuyo; Lee, Jai-Ki; Menzel, Hans-Georg; Niwa, Ohtsura; Sakai, Kazuo; Weiss, Wolfgang; Yamashita, Shunichi; Yonekura, Yoshiharu

2013-09-01

Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with 'contamination' of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of potential

Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

International Nuclear Information System (INIS)

Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

2009-10-01

The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and foreseen by for instance the

Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

Energy Technology Data Exchange (ETDEWEB)

Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

2009-10-15

The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and

Extension and update of the reliability data for fire protection equipment in German light water reactors; Ergaenzung und Aktualisierung von Zuverlaessigkeitskenngroessen fuer Brandschutzeinrichtungen in deutschen Leichtwasserreaktoren

Energy Technology Data Exchange (ETDEWEB)

Forell, Burkhard; Einarsson, Svante

2016-12-15

Within the BMUB project 3610R01370 ''Extension and update of reliability data for fire protection equipment in German light water reactors'' plant-specific as well as generic failure rates for the technical reliability of active fire protection features in German nuclear power plants have been calculated. Based on results of previous projects, in this project observation times of components were updated and extended and additional components and functions were assessed. Now, the data evaluated results from a total of six German reference plants with seven reactor units.

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

Multiple implementation of a reactor protection code in PHI2, PASCAL, and IFTRAN on the SIEMENS-330 computer

International Nuclear Information System (INIS)

Gmeiner, L.; Lemperle, W.; Voges, U.

1978-01-01

In safety related computer applications, as in the case of a reactor protection system considered here, mostly multi-computer systems are necessary for reasons of reliability and availability. The hardware structure of the protection system and the software requierements derived from it are explained. In order to study the effects of diversified programming of the three computers the protection codes were implemented in the languages IFTRAN, PASCAL, and PHI2. According to the experience gained diversified programming seems to be a proper means to prevent identical programming errors in all three computers on one hand and to detect ambiguities of the specification on the other. During all of the experiment the errors occurring were recorded in detail and at the moment are being evaluated. (orig./WB) [de

Protection method and protection device for liquid supply channel for nuclear reactor

International Nuclear Information System (INIS)

Sato, Masanori; Fujimoto, Sachiko

1998-01-01

In the present invention, thermal stresses exerted on feedwater pipelines and a feedwater nozzle portion in a LBR type reactor are reduced to improve integrity of a reactor, suppress addition of facilities and reliably reduce thermal stresses by a simple structure. A connection pipe channel is formed between the upper side of a horizontal portion of a pipeline for feeding water to a reactor pressure vessel and a vertical portion of the feedwater pipeline. A transferring pump is disposed in the midway of the connection pipe channel for sucking supernatant water in the horizontal portion and rendering it to join with water in the vertical portion. When supply of water is stagnated, high temperature water in the horizontal portion is transferred by the action of the transfer pump to the low temperature water in the vertical portion to join them. With such procedures, the water supplied to the horizontal portion in the feedwater pipeline is flown to suppress occurrence of heat-stratification phenomenon thereby enabling to reduce the temperature difference. (T.M.)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Nuclear reactor safety in the USA

International Nuclear Information System (INIS)

Vigil, J.C.

1983-01-01

Nuclear reactor safety in the USA has emphasized a defense-in-depth approach to protecting the public from reactor accidents. This approach was severely tested by the Three Mile Island accident and was found to be effective in safeguarding the public health and safety. However, the economic impact of the TMI accident was very large. Consequently, more attention is now being given to plant protection as well as public-health protection in reactor-safety studies. Sophisticated computer simulations at Los Alamos are making major contributions in this area. In terms of public risk, nuclear power plants compare favorably with other large-scale alternatives to electricity generation. Unfortunately, there is a large gulf between the real risks of nuclear power and the present public perception of these risks

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

Energy Technology Data Exchange (ETDEWEB)

Martins, Roque Hudson da Silva

2016-07-01

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

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Nuclear reactor power supply system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

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

Evaluation of the physical protection system of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Vaz, Antonio C.A.; Conti, Thadeu das N.

2013-01-01

The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

Evaluation of the physical protection system of the IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Vaz, Antonio C.A.; Conti, Thadeu das N., E-mail: acavaz@ipen.br, E-mail: tnconti@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

Foire aux questions techniques concernant la Bourse du CRDI aux ...

International Development Research Centre (IDRC) Digital Library (Canada)

Un de mes professeurs n'a pas été en mesure de remplir à temps le formulaire lui permettant de donner des références à mon sujet. Par conséquent, je n'ai pas pu soumettre ma demande avant la date et l'heure butoirs. Est-ce que le formulaire de mon professeur peut être téléversé et soumis après la date et l'heure butoirs ...

Protective head of sensors

International Nuclear Information System (INIS)

Liska, K.; Anton, P.

1987-01-01

The discovery concerns the protective heads of diagnostic assemblies of nuclear power plants for conductors of the sensors from the fuel and control parts of the said assemblies. A detailed description is presented of the design of the protective head which, as compared with the previous design, allows quick and simple assembly with reduced risk of damaging the sensors. The protective head may be used for diagnostic assemblies both in power and in research reactors and it will be used for WWER reactor assemblies. (A.K.). 3 figs

A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods

International Nuclear Information System (INIS)

Benko, Pedro Luiz

1997-01-01

A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

Guideline for Bayesian Net based Software Fault Estimation Method for Reactor Protection System

International Nuclear Information System (INIS)

Eom, Heung Seop; Park, Gee Yong; Jang, Seung Cheol

2011-01-01

The purpose of this paper is to provide a preliminary guideline for the estimation of software faults in a safety-critical software, for example, reactor protection system's software. As the fault estimation method is based on Bayesian Net which intensively uses subjective probability and informal data, it is necessary to define formal procedure of the method to minimize the variability of the results. The guideline describes assumptions, limitations and uncertainties, and the product of the fault estimation method. The procedure for conducting a software fault-estimation method is then outlined, highlighting the major tasks involved. The contents of the guideline are based on our own experience and a review of research guidelines developed for a PSA

Design strategy for control of inherently safe reactors

International Nuclear Information System (INIS)

Chisholm, G.H.

1984-01-01

Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Protection set-points lines for the reactor core and considerations about power distribution and peak factors

International Nuclear Information System (INIS)

Furieri, E.B.

1981-01-01

In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

University Reactor Matching Grants Program

International Nuclear Information System (INIS)

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-01-01

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

Reactor protection system design using application specific integrated circuits

International Nuclear Information System (INIS)

Battle, R.E.; Bryan, W.L.; Kisner, R.A.; Wilson, T.L. Jr.

1992-01-01

Implementing reactor protection systems (RPS) or other engineering safeguard systems with application specific integrated circuits (ASICs) offers significant advantages over conventional analog or software based RPSs. Conventional analog RPSs suffer from setpoints drifts and large numbers of discrete analog electronics, hardware logic, and relays which reduce reliability because of the large number of potential failures of components or interconnections. To resolve problems associated with conventional discrete RPSs and proposed software based RPS systems, a hybrid analog and digital RPS system implemented with custom ASICs is proposed. The actual design of the ASIC RPS resembles a software based RPS but the programmable software portion of each channel is implemented in a fixed digital logic design including any input variable computations. Set point drifts are zero as in proposed software systems, but the verification and validation of the computations is made easier since the computational logic an be exhaustively tested. The functionality is assured fixed because there can be no future changes to the ASIC without redesign and fabrication. Subtle error conditions caused by out of order evaluation or time dependent evaluation of system variables against protection criteria are eliminated by implementing all evaluation computations in parallel for simultaneous results. On- chip redundancy within each RPS channel and continuous self-testing of all channels provided enhanced assurance that a particular channel is available and faults are identified as soon as possible for corrective actions. The use of highly integrated ASICs to implement channel electronics rather than the use of discrete electronics greatly reduces the total number of components and interconnections in the RPS to further increase system reliability. A prototype ASIC RPS channel design and the design environment used for ASIC RPS systems design is discussed

Reactor accidents and how to protect oneself from them. Emergency measures. Schutz bei Atomunfaellen. Vorbereitet sein auf den Notfall

Energy Technology Data Exchange (ETDEWEB)

Gerosa, K

1986-01-01

The Chernobyl reactor accident has sharpened our sense of nuclear energy risks and dangers. Much as we hope it to never occur again in the future we know that there is no way of guaranteeing this to be the case. What is to be done in case of another reactor accident with radioactive radiation threatening to destroy us and the environment. Emergency measures protecting and saving our lives and health cannot be taken but with sufficient information at hand. The book informs about measures to be taken under emergency conditions. Nuclear energy and its alternatives, preventive measures, adequate nutrition and advice for pregnant women and for children are among the further subjects dealt with.

Research nuclear reactor RA - Annual report 1992

International Nuclear Information System (INIS)

Sotic, O.

1992-12-01

Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

Radiation protection/shield design

International Nuclear Information System (INIS)

Disney, R.K.

1977-01-01

Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.) [de

Determination of the protection set-points lines for the Angra-1 reactor core

International Nuclear Information System (INIS)

Furieri, E.B.

1980-03-01

In this work several thermo-hidraulic calculation were performed to obtain Protection set-points lines for the Angra-1 reactor core in order to compare with the values presented by the vendor in the FSAR. These lines are the locus of points where DNBR min = 1,3 and power = 1,18 x P nominal as a function of ΔT m and T m , the temperature difference and the average coolant temperature between hot and cold legs. A computation scheme was developed using COBRA-IIIF as a subroutine of a new main program and adding new subroutines in order to obtain the desired DNBR. The solution is obtained through a convergentce procedure using parameters estimated in a sensivity study. (author) [pt

Design of reactor protection systems for HTR plants generating electric power and process heat problems and solutions

International Nuclear Information System (INIS)

Craemer, B.; Dahm, H.; Spillekothen, H.G.

1982-06-01

The design basis of the reactor protection system (RPS) for HTR plants generating process heat and electric power is briefly described and some particularities of process heat plants are indicated. Some particularly important or exacting technical measuring positions for the RPS of a process heat HTR with 500 MWsub(th) power (PNP 500) are described and current R + D work explained. It is demonstrated that a particularly simple RPS can be realized in an HTR with modular design. (author)

Improvements in liquid metal cooled nuclear reactors

International Nuclear Information System (INIS)

Barnes, S.

1980-01-01

A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

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.

The safety of fast reactors

International Nuclear Information System (INIS)

Justin, F.

1976-01-01

A response is made to the main questions that a man in the street may arise concerning fast breeder reactors, in particular: the advantages of this line, dangerous materials contained in fast breeder reactors, containment shells protecting the environment from radiations, main studies now in progress [fr

Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux model

Energy Technology Data Exchange (ETDEWEB)

Baghban, Ghonche [Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of). Nuclear Science and Technology Research Inst.; Shayesteh, Mohsen [Imam Hussein Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics; Bahonar, Majid [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-03-15

In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.

Safety inspections to TRIGA reactors

International Nuclear Information System (INIS)

Byszewski, W.

1988-01-01

The operational safety advisory programme was created to provide useful assistance and advice from an international perspective to research reactor operators and regulators on how to enhance operational safety and radiation protection on their reactors. Safety missions cover not only the operational safety of reactors themselves, but also the safety of associated experimental loops, isotope laboratories and other experimental facilities. Safety missions are also performed on request in other Member States which are interested in receiving impartial advice and assistance in order to enhance the safety of research reactors. The results of the inspections have shown that in some countries there are problems with radiation protection practices and nuclear safety. Very often the Safety Analysis Report is not updated, regulatory supervision needs clarification and improvement, maintenance procedures should be more formalised and records and reports are not maintained properly. In many cases population density around the facility has increased affecting the validity of the original safety analysis

Protection of spent aluminum-clad research reactor fuels during extended wet storage

International Nuclear Information System (INIS)

Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

2013-01-01

Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

Evaluation of mean time between forced outage for reactor protection system using RBD and failure rate

International Nuclear Information System (INIS)

Lee, D. Y.; Park, J. H.; Hwang, I. K.; Cha, K. H.; Choi, J. K.; Lee, K. Y.; Park, J. K.

2001-01-01

The design life of nuclear power plants (NPPs) under recent construction is about fifty to sixty years. However, the duration that equipments of control systems operate without failures is at most five to ten years. Design for diversity and adequate maintenance strategy are required for NPP protection system in order to use the control equipment which has shorter life time than the design life of NPP. Fault Tree Analysis (FTA) technique, which has been applied to Probabilistics Safety Analysis (PSA), has been introduced to quantitatively evaluate the reliability of NPP I and C systems. The FTA, however, cannot properly consider the effect of maintenance. In this work, we have reviewed quantitative reliability evaluation techniques using the reliability block diagram and failure rates and applied it to the evaluation of mean time between forced outage for reactor protection system

Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

International Nuclear Information System (INIS)

Chichindaev, D.A.

2001-01-01

An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

Tasks related to increase of RA reactor exploitation and experimental potential, 01. Designing the protection chamber in the RA reactor hall for handling the radioactive experimental equipment (I-II) Part II, Vol. II

International Nuclear Information System (INIS)

Pavicevic, M.

1963-07-01

This second volume of the project for construction of the protection chamber in the RA reactor hall for handling the radioactive devices includes the technical description of the chamber, calculation of the shielding wall thickness, bottom lead plate, horizontal stability of the chamber, cost estimation, and the engineering drawings

Oregon State University TRIGA Reactor annual report

Energy Technology Data Exchange (ETDEWEB)

Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

1979-08-31

The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included.

Oregon State University TRIGA Reactor annual report

International Nuclear Information System (INIS)

Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

1979-01-01

The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included

Reports by the Parliamentary Office for scientific and technological assessments. Tuesday, May 24, 2011. Hearing on the protection of a reactor core and critical circuit; Comptes rendus de l' Office Parlementaire d'Evaluation des Choix Scientifiques et Technologiques. Mardi 24 mai 2011. Audition sur la protection du coeur et des circuits critiques d'un reacteur

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-05-15

In the context created by the Fukushima accident, members of the French Parliament, representatives of the French nuclear safety authority (ASN), of the French Institute for radiation protection and nuclear safety (IRSN), and of the CEA describe and discuss the technical aspects and mechanism of defence-in-depth of nuclear reactors (i.e. the different and successive levels of protection aimed at ensuring the reactor integrity to be maintained, even in case of failure of a critical circuit). Then, they discuss advances and researches in the field of protection of reactors. Several research programs are evoked which concern different elements of a nuclear plants such as the fuel, the reactor, loss of cooling system, and so on; these programs are based either on experiments or on simulations

RB Research nuclear reactor, Annual report for 2005

International Nuclear Information System (INIS)

Milosevic, M.; Dasic, N.; Ljubenov, V.; Pesic, M.; Nikolic, D; Jevremovic, M.; Minic, D.

2006-01-01

Report on RB reactor operation during 2005 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2005

Nuclear reactor fuel sub-assemblies

International Nuclear Information System (INIS)

Dodd, J.A.

1981-01-01

An improved fuel sub-assembly for a liquid metal cooled fast breeder reactor, is described, in which fatigue damage due to buffeting by cross-current flows is reduced and protection is provided against damage by contact with other reactor structures during loading and unloading of the sub-assembly. (U.K.)

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

Limit regulation system for pressurized water nuclear reactors

International Nuclear Information System (INIS)

Aleite, W.; Bock, H.W.

1976-01-01

Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

Reactor handbook. 2. rev. ed.

International Nuclear Information System (INIS)

Lederer, B.J.; Wildberg, D.W.

1992-01-01

On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

Radiation protection at the RA Reactor in 1994. Part 2, Annex 1, Control of the working environment, dosimetry and radiation protection

International Nuclear Information System (INIS)

Pavlovic, R.; Kalinic, S.

1994-01-01

This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. It is stated that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel [sr

ITER [International Thermonuclear Experimental Reactor] reactor building design study

International Nuclear Information System (INIS)

Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

Refurbishing the reactor protection systems of VVER-440/230 and VVER-1000/320 nuclear power plants with exclusively digital IandC systems

International Nuclear Information System (INIS)

Martin, M.

1997-01-01

The refurbishment of reactor protection systems of nuclear power plants is based on two sets of requirements: engineering aspects such as performance, qualification and licensing, as well as interfaces to other systems; and cost-benefit relationships, ease of service and maintenance as well as installation during scheduled outages. A number of WWER-440 and WWER-1000 nuclear plants have announced their intention to refurbish their protection systems. Since 1994, these plants have been placing orders with Siemens for new protection systems, including the neutron flux monitoring system utilizing the advanced system TELEPERM XS. This exclusively digital IandC system provides an excellent foundation for the remaining plant service life

Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

International Nuclear Information System (INIS)

Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

2013-01-01

RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Brunswick Steam Electric Plant, Units 1 and 2

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 Brunswick Steam Electric Plant, Units 1 and 2. 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 with time delays verified by GE, will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

Effects of ship casualties on reactor safety and marine reactor design

International Nuclear Information System (INIS)

Agena, H.H.

1978-01-01

Ship casualties such as collision, grounding, fire and explosions, leakage and partial flooding may severely impair the safety of the nuclear reactor plant and result in nuclear hazards to the crew and the environment. Engineered safeguards are being discussed for protection against such consequences: manoeuvrability, structural (passive) collision and fire protection, protection against external fires and pressure waves, after heat transmission in case of a casualty, and after heat transmission out off the sunk wreck. Existing requirements will be discussed, shortly

Plant protection system optimization studies to mitigate consequences of large breaks in the advanced neutron source reactor

International Nuclear Information System (INIS)

Khayat, M.I.; March-Leuba, J.

1993-01-01

This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the advanced neutron source (ANS) reactor. The ANS is a new basic and applied research facility based on a powerful steady-state research reactor that provides beams of neutrons for measurements and experiments in the field of material science and engineering, biology, chemistry, material analysis, and nuclear science. To achieve the high neutron fluxes for these state-of-the-art experiments, the ANS design has a very high power density core (330 MW fission with an active volume of 67.6 ell) surrounded by a large heavy-water reflector, where most neutrons are moderated. This design maximizes the number of neutrons available for experiments but results in a low heat capacity core that creates unique challenges to the design of the plant protection system

Radiation protection at the RA Reactor in 1988, Part -2, Annex 2c, Environmental radioactivity control, meteorology measurements

International Nuclear Information System (INIS)

Grsic, Z.; Zaric, M.; Adamovic, M.; Stevanovic, M.

1988-01-01

During 1988, meteorology measurements were continued as a part of the environmental control of the Vinca Institute. This report covers the period from November 1984 - November 1985. Part of the meteorology measurements and data analysis is adapted to the needs of the Institute, i.e. RA reactor and some Laboratories. The objective of these activities is forming the data base for solving everyday and special problems related to control, protection and safety of Institute environment [sr

Case study on the use of PSA methods: Assessment of technical specifications for the reactor protection system instrumentation

International Nuclear Information System (INIS)

1992-10-01

This case study presents a methodology for the probabilistic evaluation of alternative plant technical specifications regarding system surveillance frequencies and out-of-service times. The methodology is applied to the reactor protection systems of a 4 loop BWR-RESAR-3S type nuclear power plant. The effect of the statistical characteristics of the system on the relative comparison of various sets of technical specifications is examined through sensitivity studies and an uncertainty analysis. Refs, figs and tabs

Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

International Nuclear Information System (INIS)

Schweizer, Fernando Lage Araújo

2014-01-01

The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

Development of a user interface style guide for the reactor protection system cabinet operator module

International Nuclear Information System (INIS)

Lee, Hyun-Chul; Lee, Dong-Young; Lee, Jung-Woon

2004-01-01

The 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 the predefined limits. A Korean project group is developing a new digitalized RPS and the Cabinet Operator Module (COM) of the RPS is used for the RPS integrity testing and monitoring by an equipment operator. A flat panel display (FPD) with a touch screen capability is provided as a main user interface for the RPS operation. To support the RPS COM user interface design, actually the FPD screen design, we developed a user interface style guide because the system designer could not properly deal with the many general human factors design guidelines. To develop the user interface style guide, various design guideline gatherings, a walk-though with a video recorder, guideline selection with respect to user interface design elements, determination of the properties of the design elements, discussion with system designers, and a conversion of the properties into the screen design were carried out. This paper describes the process details and the findings in the course of the style guide development. (Author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

RB Research nuclear reactor, Annual report for 2006

International Nuclear Information System (INIS)

Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.

2007-01-01

Report on RB reactor operation during 2006 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains detailed data concerned with measurements performed at the RB reactor and a number of significant results obtained

Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

International Nuclear Information System (INIS)

Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

1993-12-01

Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

Generic assessment procedures for determining protective actions during a reactor accident

International Nuclear Information System (INIS)

1997-08-01

This manual provides the tools, procedures and data needed to evaluate the consequences of a nuclear accident occurring at a nuclear power plant throughout all phases of the emergency before, during and after a release of radioactive material. It is intended for use by on-site and off-site groups responsible for evaluating the accident consequences and making recommendations for the protection of the plant personnel, the emergency workers and the public. The scope of this manual is restricted to the technical assessment of radiological consequences. It does not address the emergency response infrastructure requirements, nor does it cover the emergency management aspects of accident assessment (e.g. reporting, staff qualification, shift replacement, and procedure implementation). The procedures and methods in this manual were developed based on a number of assumptions concerning the design and operation of the nuclear power plant and national practices. Therefore, this manual must be reviewed as part of the planning process to match the potential accidents, local conditions, national criteria and other unique characteristics of an area or nuclear reactor where it may be used. Refs, figs, tabs

Safety aspect of digital reactor protection system in Japan

International Nuclear Information System (INIS)

Ogiso, Zen-Ichi

1998-01-01

It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

RB Research nuclear reactor, Annual report for 2007

International Nuclear Information System (INIS)

Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.; Sipka, Dj.

2008-01-01

Report on RB reactor operation during 2007 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2007. Majority of measurement were related to spent fuel from the RA reactor, safety of transportation containers and verification of relevant computer codes

Occupational health physics at a fusion reactor

International Nuclear Information System (INIS)

Shank, K.E.; Easterly, C.E.; Shoup, R.L.

1975-01-01

Future generation of electrical power using controlled thermonuclear reactors will involve both traditional and new concerns for health protection. A review of the problems associated with exposures to tritium and magnetic fields is presented with emphasis on the occupational worker. The radiological aspects of tritium, inventories and loss rates of tritium for fusion reactors, and protection of the occupational worker are discussed. Magnetic fields in which workers may be exposed routinely and possible biological effects are also discussed

RB Research nuclear reactor, Annual report for 2004

International Nuclear Information System (INIS)

Dasic, N.; Pesic, M.; Nikolic, D; Jevremovic, M.; Eskirovic, B.

2005-02-01

Report on RB reactor operation during 2004 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. It contains data about reactor operation during previous 8 years. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation, Annex 1. contains data about heavy water degradation, and Annex 2 is the certificate about the crane bridge in the reactor hall

KS-150 reactor control

International Nuclear Information System (INIS)

Wagner, K.

1974-01-01

A thorough description is presented of the control and protection system of the Bohunice A-1 reactor. The system including auxiliary facilities was developed, manufactured and installed at the reactor by the SKODA Works, Plzen. The system parameters are listed and a brief account is also given of the development efforts and of the physical and power start-up of the A-1 nuclear power plant. (L.O.)

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Power reactor design trends

International Nuclear Information System (INIS)

Hogan, W.J.

1985-01-01

Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

A Novel Dual-Stage Hydrothermal Flow Reactor

DEFF Research Database (Denmark)

Hellstern, Henrik Christian; Becker, Jacob; Hald, Peter

2015-01-01

The dual-stage reactor is a novel continuous flow reactor with two reactors connected in series. It is designed for hydrothermal flow synthesis of nanocomposites, in which a single particle consists of multiple materials. The secondary material may protect the core nanoparticle from oxidation....... The dual-stage reactor combines the ability to produce advanced materials with an upscaled capacity in excess of 10 g/hour (dry mass). TiO2 was synthesized in the primary reactor and reproduced previous results. The dual-stage capability was succesfully demonstrated with a series of nanocomposites incl. Ti...

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

International Nuclear Information System (INIS)

Souza Mendes, J.E. de

1987-01-01

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

Elements on reactor control

International Nuclear Information System (INIS)

Bruna, G.B.

1998-01-01

In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

Advanced spheromak fusion reactor

International Nuclear Information System (INIS)

Fowler, T.K.

1996-01-01

The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

Design study of 'HIBLIC-I' reactor cavity

International Nuclear Information System (INIS)

Fujiie, Y.

1984-01-01

A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

EPRTM Reactor neutron instrumentation

International Nuclear Information System (INIS)

Pfeiffer, Maxime; SALA, Stephanie

2013-06-01

The core safety during operation is linked, in particular, to the respect of criteria related to the heat generated in fuel rods and to the heat exchange between the rods and the coolant. This local power information is linked to the power distribution in the core. In order to evaluate the core power distribution, the EPR TM reactor relies on several types of neutron detectors: - ionization chambers located outside the vessel and used for protection and monitoring - a fixed in-core instrumentation based on Cobalt Self Powered Neutron Detectors used for protection and monitoring - a mobile reference in-core instrumentation based on Vanadium aero-balls This document provides a description of this instrumentation and its use in core protection, limitation, monitoring and control functions. In particular, a description of the detectors and the principles of their signal generation is supplied as well as the description of the treatments related to these detectors in the EPR TM reactor I and C systems (including periodical calibration). (authors)

Programming Guidelines for FBD Programs in Reactor Protection System Software

International Nuclear Information System (INIS)

Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom; Lee, Jang Su

2014-01-01

Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows

Programming Guidelines for FBD Programs in Reactor Protection System Software

Energy Technology Data Exchange (ETDEWEB)

Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom [Division of Computer Science and Engineering College of Information and Communication, Konkuk University, Seoul (Korea, Republic of); Lee, Jang Su [Man-Machine Interface System team Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows.

Introduction to symposium 'radiation protection at nuclear facilities'

International Nuclear Information System (INIS)

Stricker, L.

1996-01-01

An introduction to the symposium 'radiation protection of nuclear facilities' on Wednesday, April 17, 1996 in Vienna has been given. The number of operating reactors and the total collective dose per reactor in OECD countries has been discussed. The evolution of the total collective dose associated with the replacement of steam generators at nuclear power reactors from 1979 to 1995 is presented. The background and culture of radiation protection, regulatory aspects, strategic formulation, plan management policy and organization responsibilities are discussed generally. (Suda)

Behavior of the future LHC magnet protection diodes irradiated in a nuclear reactor at 4.6 K with intermediate annealing

International Nuclear Information System (INIS)

Berland, V.; Hagedorn, D.; Gerstenberg, H.

1996-01-01

In the framework of the LHC project at CERN, the effects of radiation on the electrical characteristics of epitaxial diodes for superconducting magnet protection were studied. The diodes were exposed to an irradiation dose up to 50 kGy and a neutron fluence of 10 15 n/cm 2 with intermediate thermal annealing each 10 kGy dose steps in the Technical University of Munich reactor at 4.6 K

RB research nuclear reactor, Annual report for 1984, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

1984-01-01

The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

Fundamentals of boiling water reactor systems

International Nuclear Information System (INIS)

Mattern, J.

1976-01-01

The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

Oscillating liquid flow ICF Reactor

International Nuclear Information System (INIS)

Petzoldt, R.W.

1990-01-01

Oscillating liquid flow in a falling molten salt inertial confinement fusion reactor is predicted to rapidly clear driver beam paths of residual liquid droplets. Oscillating flow will also provide adequate neutron and x-ray protection for the reactor structure with a short (2-m) fall distance permitting an 8 Hz repetition rate. A reactor chamber configuration is presented with specific features to clear the entire heavy-ion beam path of splashed molten salt. The structural components, including the structure between beam ports, are shielded. 3 refs., 12 figs

Experimental Breeder Reactor I Preservation Plan

Energy Technology Data Exchange (ETDEWEB)

Julie Braun

2006-10-01

Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

Description of the Triton reactor

International Nuclear Information System (INIS)

1967-09-01

The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965 [fr

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

1994-04-01

Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

International Nuclear Information System (INIS)

Korsah, K.; Clark, R.L.; Wood, R.T.

1994-04-01

Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I ampersand C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems' environmental qualification and functional reliability. To bound the problem of new I ampersand C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I ampersand C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I ampersand C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software

Reactor control device

International Nuclear Information System (INIS)

Araki, Takao; Inoue, Toyokazu.

1981-01-01

Purpose: To protect the reactor floor by alleviating the shock imparted to the reactor floor by a dropped control rod when a wire rope accidentally breaks. Constitution: A control rod is hung by wire rope from a control rod drive, and shock absorbers are mounted at the upper and lower portions of the control rod. The outer diameter of the upper shock absorber is made larger than the inner diameter of a control rod inserting hole formed in the reactor core. If the control rod drops, the upper absorber is stopped at the upper tapered portion of the inserting hole. Thus, the dropping energy of the control rod can be sufficiently absorbed by the upper and lower shock absorbers. (Kamimura, M.)

Safety equipment in a reactor

International Nuclear Information System (INIS)

Shiratori, Hirozo; Ishiyama, Satoshi; Ugawa, Yukio.

1976-01-01

Object: To safely retain, even if fuel should be molten and flown through the bottom of a container in a reactor, the molten fuel to remove heat generation of the fuel to prevent occurrence of a critical trouble. Structure: A reactor container housing a core and coolant has thereunder a separation dome in a central portion thereof and a partitioning plate coaxially and circularly disposed in the periphery of the separation dome, with a tray formed of magnesium oxide being disposed. Further, a cooling path system is provided so as to surround the tray. The cooling path system and the reactor container are surrounded and protected by a reactor wall provided with heat insulating refractory bricks, a coolant pouring system extends through the reactor wall, and the coolant is supplied to the tray. (Furukawa, Y.)

Safety study of fire protection for nuclear fuel cycle facility

International Nuclear Information System (INIS)

2013-01-01

Insufficiencies in the fire protection system of the nuclear reactor facilities were pointed out when the fire occurred due to the Niigata prefecture-Chuetsu-oki Earthquake in July, 2007. This prompted the revision of the fire protection safety examination guideline for nuclear reactors as well as commercial guidelines. The commercial guidelines have been endorsed by the regulatory body. Now commercial fire protection standards for nuclear facilities such as the design guideline and the management guideline for protecting fire in the Light Water Reactors (LWRs) are available, however, those to apply to the nuclear fuel cycle facilities such as mixed oxide fuel fabrication facility (MFFF) have not been established. For the improvement of fire protection system of the nuclear fuel cycle facilities, the development of a standard for the fire protection, corresponding to the commercial standard for LWRs were required. Thus, Japan Nuclear Energy Safety Organization (JNES) formulated a fire protection guidelines for nuclear fuel cycle facilities as a standard relevant to the fire protection of the nuclear fuel cycle facilities considering functions specific to the nuclear fuel cycle facilities. In formulating the guidelines, investigation has been conduced on the commercial guidelines for nuclear reactors in Japan and the standards relevant to the fire protection of nuclear facilities in USA and other countries as well as non-nuclear industrial fire protection standards. The guideline consists of two parts; Equipments and Management, as the commercial guidances of the nuclear reactor. In addition, the acquisition of fire evaluation data for a components (an electric cabinet, cable, oil etc.) targeted for spread of fire and the evaluation model of fire source were continued for the fire hazard analysis (FHA). (author)

EPR by Areva. EPR the 1600+ MWe reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system.

EPR by Areva. EPR the 1600+ MWe reactor

International Nuclear Information System (INIS)

2008-01-01

This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system

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)

Aircraft-crash-protected roof design for the European SBWR

International Nuclear Information System (INIS)

Posta, B.A.; Kadar, I.; Rao, A.S.

1995-01-01

The European utility requirement document (EURD) places significant emphasis on aircraft crash protection of the reactor building - Alternative concepts were evaluated for protecting the dry-well head and the fuel pool from the effect of the spalling concrete for the General Electric Company's European simplified boiling water reactor (ESBWR) designs

Research nuclear reactor RA - Annual Report 1990 with the comparative evaluation for the period 1986-1990

International Nuclear Information System (INIS)

Sotic, O.

1990-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1990, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Agriculture protected against nuclear power hazards

International Nuclear Information System (INIS)

Warner, J.

1982-01-01

Near the Koeberg-1 reactor agriculture is being protected against nuclear power hazards. Radioactivity is frequently tested in e.g. milk and soils. This article discusses the tests and the measurements of radioactivity in the vicinity of the Koeberg-1 reactor

Preliminary Validation and Verification Plan for CAREM Reactor Protection System; Modelo de Plan Preliminar de Validacion y Verificacion para el Sistema de Proteccion del Reactor CAREM

Energy Technology Data Exchange (ETDEWEB)

Fittipaldi, Ana; Felix, Maciel [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

2000-07-01

The purpose of this paper, is to present a preliminary validation and verification plan for a particular architecture proposed for the CAREM reactor protection system with software modules (computer based system).These software modules can be either own design systems or systems based in commercial modules such as programmable logic controllers (PLC) redundant of last generation.During this study, it was seen that this plan can also be used as a validation and verification plan of commercial products (COTS, commercial off the shelf) and/or smart transmitters.The software life cycle proposed and its features are presented, and also the advantages of the preliminary validation and verification plan.

Official announcement of the directive on protection of nuclear power plant equipped with LWR-type reactors from human intrusion or other interference by third parties. Announcement of BMU (German Federal Ministry Environment), of 6 Dec. 1995 - RS I 3 13151 - 6/14

International Nuclear Information System (INIS)

1996-01-01

An operating permit for a nuclear power plant is to be granted only if the applicant and facility operator presents evidence guaranteeing the legally required physical protection and other security measures for protection from human instrusion and other type of interference. As a basis for review and licensing, the competent authorities in 1987 have issued a directive specifying the requirements to be met for physical protection of nuclear power plant equipped with PWR-type reactors, and in 1994 followed a second, analogous directive relating to nuclear power plant with BWR-type reactors. The directive now announced for physical protection of nuclear power plant equipped with LWR-type reactors combines and replaces the two former ones, and from the date of the announcement is the only applicable directive. The text of the directive is not reproduced for reasons of secrecy protection. (orig./CB) [de

Optical design considerations for laser fusion reactors

International Nuclear Information System (INIS)

Monsler, M.J.; Maniscalco, J.A.

1977-09-01

The plan for the development of commercial inertial confinement fusion (ICF) power plants is discussed, emphasizing the utilization of the unique features of laser fusion to arrive at conceptual designs for reactors and optical systems which minimize the need for advanced materials and techniques requiring expensive test facilities. A conceptual design for a liquid lithium fall reactor is described which successfully deals with the hostile x-ray and neutron environment and promises to last the 30 year plant lifetime. Schemes for protecting the final focusing optics are described which are both compatible with this reactor system and show promise of surviving a full year in order to minimize costly downtime. Damage mechanisms and protection techniques are discussed, and a recommendation is made for a high f-number metal mirror final focusing system

Technical evaluation report on the monitoring of electric power to the reactor protection system for the Nine Mile Point Nuclear Station, Unit 1 (Docket No. 50-220)

International Nuclear Information System (INIS)

Selan, J.C.

1984-01-01

This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Nine Mile Point Nuclear Station, Unit 1. 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

RB research nuclear reactor, Annual report for 1983, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

1983-01-01

The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

Common cause analysis of the TREAT upgrade reactor protection system

Energy Technology Data Exchange (ETDEWEB)

Page, R.J.; Kamis, G.J.; Marbach, R.A.; Mueller, C.J.

1984-09-01

A triply redundant reactor scram system (RSS) has been designed for the upgraded TREAT facility. The independent failures reliability goal for the RSS is <10/sup -9/ failures per demand. An independent failures analysis indicated that this goal would be met. In addition, however, recognizing that in heavily redundant systems common-cause failures dominate, a common cause analysis of the TREAT upgrade RSS was done. The objective was to identify those common-cause initiators which could affect the functioning of the RSS, and to subsequently modify the design of the RSS so that the effect was minimized. A number of common-cause initiators were identified which were capable of defeating the triple redundancy feature of the reactor scram system. By means of a systematic analysis of the effect these initiators could have on the system, it was possible to identify seven necessary design and procedural modifications that would greatly reduce the probability of the reactor being run while the RSS was in a faulted condition.

Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

International Nuclear Information System (INIS)

Ramanathan, Lalgudi V.

2009-01-01

A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Operation and maintenance of the RB reactor, Annual report for 1977

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1977-01-01

The annual report for 1977 includes the following: utilization of the RB reactor; new regulations and instructions for reactor operation; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Five annexes are concerned with: testing the properties of preamplifiers for linear and logarithmic experimental channels; properties of the neutron converter; maintenance of the reactor equipment; purchase of new equipment; and the program for training reactor operators

Human Factors Support in the Design and Evaluation of the Reactor Protection System Cabinet Operator Module

International Nuclear Information System (INIS)

Lee, Hyun Chul; Lee, Jung Woon

2005-01-01

A Korean project group, KNICS, is developing a new digitalized reactor protection system (RPS) and the developed system will be packaged into a cabinet with several racks. The cabinet of the RPS is used for the RPS function testing and monitoring by maintenance operators and is equipped with a flat panel display (FPD) with a touch screen capability as a main user interface for the RPS operation. This paper describes the human factors activities involved in the development process of the RPS: conceptual design, design guidance, and evaluation. The activities include a functional requirements analysis and task analysis, user interface style guide for the RPS cabinet operator module (COM), and a human factors evaluation through an experiment and questionnaires

RB research nuclear reactor, Annual report for 1989, I - III

International Nuclear Information System (INIS)

Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

1989-12-01

This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

RB Research nuclear reactor, Annual report for 1996, I-IV

International Nuclear Information System (INIS)

Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

1996-12-01

Report on RB reactor operation during 1996 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a list of publications resulting from experiments done at the RB reactor

RB Research nuclear reactor, Annual report for 1995, I-IV

International Nuclear Information System (INIS)

Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

1995-12-01

Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor

Anticorrosional protection in nuclear power station objects

International Nuclear Information System (INIS)

Czarnocki, A.; Kwiatkowski, A.

1976-01-01

The distribution and qualities of chemical protection and demands concerning preparation of the bottom for protecting coats in nuclear power station objects are discussed. The solutions of protections applied abroad and in the objects of ''MARIA'' reactor are presented. (author)

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)

Self-protection of FLIP fuel: Experimental measurements

International Nuclear Information System (INIS)

Johnson, A.G.; Ringle, J.C.; Anderson, T.V.; Dodd, B.; Bennett, S.; Carpenter, W.T.

1980-01-01

During the last year and one-half non-power reactor licensees of the Nuclear Regulatory Commission (NRC) have been repeatedly alerted to the fact that the Commission planned to tighten physical security regulations for their class of reactors. Essentially all of the new NRC physical security requirements for these reactors have been based on the type, quantity and enrichment of the special nuclear materials (e.g. enriched uranium and plutonium) possessed by the licensees. While each licensee's security classification will be evaluated individually, it is now apparent that non-power reactor licensees using or storing formula quantities of special nuclear material not meeting the 100 rem per hour at three feet self-protection exemption (e.g. 5 kilograms or more of non-self-protected fuel enriched to 20% or more with U-235) will be required to meet extremely stringent NRC physical security requirements patterned after the new Physical Protection Upgrade Rule. Implementation of these much stricter security requirements poses many economic and operational difficulties for university research reactors, to the point where facility closure could be the only alternative. TRIGA reactors utilizing a full FLIP fueled core, such as the Oregon State University TRIGA (OSTR), qualify for the highest physical protection category unless fuel can be maintained at self-protecting radiation levels. In order to demonstrate that OSTR fuel could be consistently kept above the 100 rem per hour threshold, a computer program was written which predicts the gamma radiation levels from an irradiated FLIP fuel element at 1, 2 and 3 feet in air and water. Furthermore, in order to verify the accuracy of the computer program, actual measurements of irradiated fuel elements were made at 3 feet in air and at 1, 2 and 3 feet in water, and the results compared very favorably to the predictions. The results of specific measurements, the instrumentation used and its calibration, the personnel doses

Reactors with pressure vessel in pre-stressed concrete

International Nuclear Information System (INIS)

Devillers, Christian; Lafore, Pierre

1964-12-01

After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Science.gov (United States)

2013-12-09

[email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...: timothy.kolb@nrc.gov . SUPPLEMENTARY INFORMATION: I. Accessing Information and Submitting Comments A...

Turbine protecting device in a BWR type nuclear power plant

International Nuclear Information System (INIS)

Kasuga, Hajime; Oka, Yoko.

1984-01-01

Purpose: To prevent highly humid steams from flowing into the turbine upon abnormal reduction in the reactor water level in order to ensure the turbine soundness, as well as in order to trip the turbine with no undesired effect on the reactor. Constitution: A protection device comprising a judging device and a timer are disposed in a BWR type reactor, in order to control a water level signal from a reactor water level gage. If the reactor water level is reduced during rated power operation, steams are kept to be generated due to decay heat although reactor is scramed. When a signal from the reactor water level detector is inputted to the protection device, a trip signal is outputted by way of a judging device after 15 second by means of the timer, when the main steam check valve is closed to trip the turbine. With this delay of time, abrupt increase in the pressure of the reactor due to sudden shutdown can be prevented. (Nakamoto, H)

Recent U.S. reactor operating experience

International Nuclear Information System (INIS)

Stello, V. Jr.

1977-01-01

A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

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)

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

A feasibility study on long-life reduced-moderation water reactor with highly protected Pu breeding by doping with minor actinides

International Nuclear Information System (INIS)

Hamase, Erina; Damian, Frederic; Poinot-salanon, Christine; Saito, Masaki; Sagara, Hiroshi; Han, Chi Young

2013-01-01

Highlights: ► The present paper is focused on a reduced-moderation water reactor. ►237 Np or 241 Am was doped into the inner blanket. ►238 Pu transmuted from MA works as a fissionable nuclide. ► It also contributes to increasing the proliferation resistance of Pu. ► Long-life core and highly protected Pu breeding were simultaneously achieved. - Abstract: The present paper is focused on the feasibility of reduced-moderation water reactor (RMWR) which could simultaneously achieve the extension of core life-time, and highly protected plutonium (Pu) breeding performance with short compound system doubling time (CSDT) by neptunium-237 ( 237 Np) or americium-241 ( 241 Am) doping of the inner blanket of the RMWR. As a preliminary analysis of the RMWR, a simplified fuel pin configuration was analyzed. In case of 60% doping of 237 Np and 241 Am in the inner blanket, the maximum available effective full power days (EFPDs) were much longer and were about 15,100 and 14,200 EFPDs, respectively. For the proliferation resistance of Pu produced in the blanket, the total Pu in the inner and lower/upper blanket was below the practically unusable criterion for an explosive device proposed by Pellaud, and was technically unfeasible for high-technology hypothetical nuclear explosive devices (HNEDs) proposed by Kessler and Kimura. The protected Pu breeding performance was analyzed and in case of 60% doping of 237 Np and 241 Am, CSDT was short by approximately 40 and 50 years. Therefore, the feasibility of RMWR was shown which could simultaneously achieve the extension of core life-time and have highly protected Pu breeding performance with short CSDT by doping the inner blanket with minor actinides (MAs). The objective of this study was thus to evaluate the performance of the concept with respect to the core life-time, proliferation resistance of Pu and CSDT. The technological feasibility of such core concept will have to be evaluated by further dedicated analyses

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

A swivelling transfer device for nuclear reactors

International Nuclear Information System (INIS)

Allain, Albert; Mulot, Pierre; Filloleau, Etienne

1974-01-01

The invention relates to a swivelling transfer device for fuel-assemblies. According to the invention, the device comprises, within a protective enclosure, a swivelling system comprising two sets of rails rotatable about an axis and so arranged that the lower and thereof penetrates into the extensions of the extremities of ramps dipped into the reactor and into a storage enclosure. This can apply to the transfer of nuclear reactor fuel assemblies, in particular for reactors of the molten sodium fast neutron type [fr

Integral isolation valve systems for loss of coolant accident protection

Science.gov (United States)

Kanuch, David J.; DiFilipo, Paul P.

2018-03-20

A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

Research Project 'RB research nuclear reactor' (operation and maintenance), Final report

International Nuclear Information System (INIS)

1985-01-01

This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training

Operation and maintenance of the RB reactor, Annual report for 1978

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1978-01-01

The annual report for 1978 includes the following: utilization of the RB reactor; producing the new safety report; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Four annexes to this report are concerned with: operation of the reactor at higher power levels; performance of the instrumentation, radiation doses during operation; gamma radiation doses after reactor shutdown; properties of the neutron converter (optimization of the rector-converter coupling; maintenance of the reactor equipment; purchase of new equipment

Specific features of using programmed hardware in emergency protection systems of NPP

International Nuclear Information System (INIS)

Konoplev, N.P.; Pogorelov, I.V.; Kirsanov, A.V.

2005-01-01

The specific features of using the microprocessor-based technical means and programmed logic integrated circuits (PLIC) in WWER reactor protection systems are considered. The basic causes of occurring the failures in programmed technical means are analyzed. The features of using PLIC in reactor protection systems are discussed. The methods permitting to decrease the probability occurring the common mode failure of programmed technical means are proposed. It is concluded that at creating of reactor protection systems with the use of microprocessors and PLIC it is necessary to apply the principle of the functional diversity [ru

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)

Researching and improving the reliability of reactor protection system of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Jiang Zuyue; Sheng Jiannan

1997-01-01

Due to the original design defects of the Reactor Protection System (RPS) of Qinshan Nuclear Power Plant, this system has brought about a number of reactor shutdown accidents and Engineered Safety Features (ESF) mis-activation events which have seriously endangered safe and steady operation of the nuclear power plant. So over three years have been spent on research on the reform of the original design on the premise that the general wiring of the system should remain the same and that the system size should also remain small to be contained in the original cabinets. The following improvements were made: (1) Increase the system's anti-disturbance capability. The system's zero volt bus floating designs were modified to surmount the disturbance resulting from the bad isolation performance of impedance-isolated amplifier; Double grounds have been added to logical modules to surmount the disturbance resulting from zero volt floating bus during the replacement of single module with two connectors; The opto-coupling circuit in its oscillation input stage of Engineered Safety Features have been improved to increase its reliability. (2)Modify to output activation part of the system. The new type of output relays were selected and the relay activation circuits were redesigned in which switcher activation mode is used instead of amplifier activation mode so as to increase the reliability of relay operation and reduce the power consumption; CMOS buffer gates in the input and output stage of the circuit were used to match TTL circuits to CMOS circuits of the system

Research nuclear reactor RA, Annual Report 2001

International Nuclear Information System (INIS)

Sotic, O.

2002-01-01

During 2001, activities at the RA research nuclear reactor in were performed according to the Contract about financing of the RA reactor for the period January-December 2001, signed by the Ministry of Science, technology and development of the Republic of Serbia. RA reactor was not operated since shutdown in August 1984. Although, the most of the planned reconstruction activities were finished until 1991, the most important, which was concerned with exchange of the reactor instrumentation, financed by the IAEA, was interrupted due to international sanctions imposed on the country. Since 1992, all the renewal and reconstruction activities were ceased. Continuous aging and degradation of the equipment and facilities demand decision making about the future status of the Ra reactor. Until this decision is made it is an obligation to maintain control and maintenance of ventilation system, power supply, internal transportation system, spent fuel storage, hot cells, electronic fuel surveillance system, and part of the stationary dosimetry system. In 2001, apart from the mentioned activities, actions were undertaken related to maintenance of the reactor building and installations. The most important tasks fulfilled were: protection of the roof of the ventilation system building, purchase and installing the fire protection system and twelve new battery cells in the reactor building. There were no actions concerned with improvement of the conditions for intermediate spent fuel storage. With the support of IAEA, actions were initiated for possible transport of the spent fuel tu Russia. At the end of 2001, preparations were started for possible future decommissioning of the RA reactor. After, renewal of the membership of our country in the IAEA, the Government of Yugoslavia has declared its attitude about the intention of RA reactor decommissioning at the General Conference in September 2001 [sr

Operation and maintenance of the RB reactor, Annual report for 1980

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.; Markovic, H.; Zivkovic, B.; Gogdanovic, M.; Petronijevic, M.

1980-12-01

This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 9 Annexes as follows: Utilization of the RB reactor from 1976 - 1980; program of reactor utilization from 1981-1985; contents of the RB reactor safety report; maintenance of the reactor components and equipment in 1980; verification of reactor reliability after the earthquake (May 18 1980); refurbishment of equipment in 1980, and purchasing new equipment from 1981-1985; review of radiation doses in the reactor building and exposure doses for the reactor staff; personnel data and financial data

Practical application of computer program Panthere for workers' radiation protection

International Nuclear Information System (INIS)

Carlier, Pierre; Michoux, Xavier; Lereculey, Clement

2014-01-01

The civil engineering operations to strengthen the raft of Fessenheim's nuclear plant were carried out by EDF. This technical modification has two principles objectives: 1) to increase the thickness of the reactor pit's concrete and 2) to create a new spreading area for corium (by creating a penetration through the wall of the reactor pit). Behind the complex technical operations two radioprotection issues were studied using the computer program 'PANTHERE': 1) Workers' radiation protection during the execution of the work (because of high dose rates in the reactor pit) and 2) operators' radiation protection after the execution of the work. Results contributed to decrease personal and collective dosimetry of operations and to model and design a biological shield to protect workers during Fessenheim reactor operation. (authors)

Nuclear reactor can detonate like an atomic bomb: yes or no

International Nuclear Information System (INIS)

Martin, D.

1974-01-01

The fission process in a nuclear weapon and in power reactor are compared. The operation of a power reactor is described and the various protective systems and shielding devices are mentioned. It is shown that it is not possible for a nuclear explosion to occur in a power reactor

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)

Impacts on power reactor health physics programs

International Nuclear Information System (INIS)

Meyer, B.A.

1991-01-01

The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

EPR (European Pressurized Reactor)

International Nuclear Information System (INIS)

2015-01-01

This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-07

At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

International Nuclear Information System (INIS)

1994-01-01

At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

Local AREA networks in advanced nuclear reactors

International Nuclear Information System (INIS)

Bicknell, J.; Keats, A.B.

1984-01-01

The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

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

Novelties in design and construction of the advanced reactors

International Nuclear Information System (INIS)

Acosta Ezcurra, T.; Garcia Rodriguez, B.M.

1996-01-01

The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

Irradiation of quench protection diodes at cryogenic temperatures in a nuclear research reactor

International Nuclear Information System (INIS)

Hagedorn, D.; Schoenbacher, H.; Gerstenberg, H.

1996-01-01

Within the framework of the Large Hadron Collider (LHC) R ampersand D programme, CERN and the Department of Physics E21 of the Technical University Munich have established a collaboration to carry out irradiation experiments at liquid helium and liquid nitrogen temperatures on epitaxial diodes for the superconducting magnet protection. Small diode samples of 10 mm wafer diameter from two different manufacturers were submitted to doses of up 50 kGy and neutron fluences up to 1015 n/cm 2 and the degradation of the electrical characteristics was measured versus dose. During irradiation the diodes were submitted to current pulse annealing and after irradiation to thermal annealing. After exposure some diodes show a degradation in forward voltage drop of up to 600 % which, however, can be reduced to about 15 % - 20 % by thermal annealing. The degradation at liquid helium temperature is very similar to the degradation at liquid nitrogen temperature. These degradations of electrical characteristics during the short term irradiation in a nuclear reactor are compared with degradations during long term irradiation in an accelerator environment at liquid nitrogen temperature

Installation and commissioning of operation nuclear power plant reactor protection system modernization project

International Nuclear Information System (INIS)

Lu Weiwei

2010-01-01

Qinshan Nuclear Power Plant is the first nuclear power plant in mainland China; it is also the first one which realizes the modernization of analog technology based Reactor Protection System in the operation nuclear power plant of China. The implementation schedule is the shortest one which use same digital technology platform (TELEPERM XS of AREVA NP) to modifying the safety class I and C system in the world, the whole project spent 28 months from equipment contract signed to putting system into operation. It open up a era for operation nuclear power plant using mature digital technology to make safety class I and C system modernization in China. The important practical significance of this successful project is very obvious. This article focus on two important project stage--equipment installation and system commissioning, it is based on a large number of engineering implementation fact, it covers the problems and solutions happened during the installation and commission. The purpose of the article is to share the experience and lessons of safety I and C system modernization for other operation nuclear power plant. (authors)

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

A Research on Seamless Platform Change of Reactor Protection System From PLC to FPGA

Energy Technology Data Exchange (ETDEWEB)

Yoo, Junbeom; Lee, Jonghoon [Konkuk Univ., Seoul (Korea, Republic of); Lee, Jangsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-08-15

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

A Research on Seamless Platform Change of Reactor Protection System From PLC to FPGA

International Nuclear Information System (INIS)

Yoo, Junbeom; Lee, Jonghoon; Lee, Jangsoo

2013-01-01

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea

A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA

Directory of Open Access Journals (Sweden)

JUNBEOM YOO

2013-08-01

Full Text Available The PLC (Programmable Logic Controller has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems. Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array. Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

Nuclear reactor safety: physics and engineering aspects

International Nuclear Information System (INIS)

Kinchin, G.H.

1982-01-01

In order to carry out the sort of probabilistic analysis referred to by Farmer (Contemp. Phys.; 22:349(1981)), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in sections dealing with the neutron chain reaction, the removal of heat from the reactor, material problems, reliability of protective systems and a number of specific topics of particular interest from the point of view of physics or engineering. (author)

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

Analysis of the radiometric survey during the Argonauta reactor operation

International Nuclear Information System (INIS)

Oliveira, Eara de S.L.; Cardozo, Katia K.M.; Silva, Joao Carlos P.; Santos, Joao Regis dos

2013-01-01

The Argonaut reactor at the Institute of Nuclear Engineering-IEN/CNEN, operates normally, the powers between 1.7 and 340 W on neutrongraphy procedures, production of radionuclides and experimental reactor physics lessons to postgraduate courses. The doses from neutrons and gamma radiation are measured when the reactor is critical, inside the reactor hall and surrounding regions. A study of the data obtained was performed to evaluate the daily need of this survey in the reactor hall. Taking into account the principle ALARA, which aims to optimize and minimize the dose received by the individual, we propose, in this work, through an analysis of the acquired data in occupational radiometric surveys, a reformulation of the area monitoring routine practiced by the team of radiological protection of the Institute of Nuclear Engineering - IEN/CNEN-RJ, whereas other monitoring routines regarding the radiological protection are also applied in the routine of the reactor. The operations under review occurred with the reactor operating 340 W power at intervals of 60, 120 and 180 minutes, in monitoring points in controlled areas, supervised and free. The results showed significant dose values in the output of the J-Channel 9 when the operation occurs with this open. With 180 minutes of operation, the measured values of dose rate were lower than the values at 60 min and 120 operations min. At the point in the supervised area, offsite to the reactor hall, situated in the direction of the J-Channel 9, the value reduces more than 14% in any operating time in relation to the dose rate measured at the point opposite the canal. There is a 50% reduction in the dose rates for operations with and J-9 closed. The results suggest a new frequency of radiometric survey whose mode of operation is maintained in similar conditions, since combined with other relevant practices of radiation protection

Lessons from feedback of safety operating experience for reactor physics

International Nuclear Information System (INIS)

Suchomel, J.; Rapavy, S.

1999-01-01

Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

Ground test facility for nuclear testing of space reactor subsystems

International Nuclear Information System (INIS)

Quapp, W.J.; Watts, K.D.

1985-01-01

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

Thermal Reactor Safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Thermal Reactor Safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Science.gov (United States)

2010-01-01

... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF...

Radiation protection at the RA Reactor in 1985, Part -2, Annex 1, Radioactivity control of working environment, dosimetry

International Nuclear Information System (INIS)

Ninkovic, M.; Bjelanovic, J.; Minincic, Z.; Komatina, R.; Raicevic, J.

1985-01-01

This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted to 8.2 mSV during past 10 months. Individual exposures for 7/10 of the personnel were less than 1/10 of the annual permissible exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. The last part analyzes accidents occurred at the reactor during 1985. It was found that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel [sr

Operation and maintenance of the RB reactor, Annual report for 1979

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.; Petronijevic, M.

1979-01-01

The annual report for 1979 includes the following: utilization of the RB reactor; accident analysis; dosimetry and radiation protection; description of the reactor equipment status; reactor staff; financial data. Six Annexes to this report are concerned with: scientific program orientation; producing the new safety report; maintenance of the reactor components and equipment; purchase of new equipment; financial report; and program for training the staff of the Krsko NPP

Separation review program for reactor protection system and engineered safeguard systems in a nuclear power plant

International Nuclear Information System (INIS)

Lamb, F.J.; Walrod, B.E.

1980-01-01

This review program is utilized during the design of a nuclear power plant to insure separation between interdiscipline design for the Reactor Protection System (RPS) and Engineered Safeguard Systems (ESS). Color coded transparent drawings of the RPS and ESS are produced by each discipline. The separation is then reviewed by overlaying drawings of different disciplines on a light table. When this inspection shows that RPS or ESS elements have less than the established minimum separation, an analysis is performed to determine what, if any, design revision is necessary to insure proper separation. ''Hazard'' drawings are also made for determination of each type of potential hazard in each area of the plant. The review is a continuing process as the design progresses and is revised by any discipline. 5 refs

Project RA Research nuclear reactor - Annual report 1993 with comparative review for the period 1991 - 1993

International Nuclear Information System (INIS)

1993-12-01

Research reactor RA Annual report for year 1993 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. The ninth separate annex deals with the feasibility of RA reactor applications. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

Technical improvement for the output drive unit of the reactor protection system in QNPP

International Nuclear Information System (INIS)

Jiang Zuyue

1995-11-01

For improving the reliability of the output drive unit of the reactor protection system in Qinshan NPP, the former design of this part was improved and researched on the problem appeared during the commissioning and operation under the conditions of narrow process space of cabinets and unchanged overall arrangement: (1) The output relay modules was redesigned to unify the relay specification to improve the versatility, and also to improve the pin's contact by means of welding them directly on the printed circuit boards and to make the modules detachable by connectors instead of previously non-detachable. Th modules were connected in series by both power supply line and ground line which were finally connected at same point respectively, so that other protection signals can still be output correctly when a single module is removed. (2) The relay drive circuit was also redesigned for working in on-off state instead of in amplification to minimize the power consumption. On the other hand, the CMOS buffers were taken to couple the CMOS circuits to the TTL circuits. The actuating time for the new shutdown relay was decreased from the former 35 ms to 5 ms, the actuating time for the engineered safety feature drive signal relay was decreased from 10 ms to 6 ms after the above-mentioned improvements, the reliability of the RPS is remarkably improved and a great economic benefit is obtained. (4 refs., 3 figs., 2 tabs.)

Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

International Nuclear Information System (INIS)

1986-11-01

This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

Physical and technical aspects of lead cooled fast reactors safety

International Nuclear Information System (INIS)

Orlov, V.V.; Smirnov, V.S.; Filin, A.I.

2001-01-01

The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

Decommissioning of Salaspils nuclear reactor

International Nuclear Information System (INIS)

Abramenkovs, A.; Malnachs, J.; Popelis, A.

2002-01-01

In May 1995, the Latvian Government decided to shut down the Research Reactor Salaspils (SRR) and to dispense with nuclear energy in future. The reactor has been out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH from 1998-1999. he Latvian Government decided on 26 October 1999 to start the direct dismantling to 'green field' in 2001. The results of decommissioning and dismantling performed in 1999-2001 are presented and discussed. The main efforts were devoted to collecting and conditioning 'historical' radioactive waste from different storages outside and inside the reactor hall. All radioactive material more than 20 tons were conditioned in concrete containers for disposal in the radioactive waste depository 'Radons' in the Baldone site. Personal protective and radiation measurement equipment was upgraded significantly. All non-radioactive equipment and material outside the reactor buildings were free-released and dismantled for reuse or conventional disposal. Weakly contaminated material from the reactor hall was collected and removed for free-release measurements. The technology of dismantling of the reactor's systems, i.e. second cooling circuit, zero power reactors and equipment, is discussed in the paper. (author)

Qualification of the Darwin code for the studies of the fuel cycle relative to the boiling water reactors; Qualification du formulaire Darwin pour les etudes du cycle du combustible pour les reacteurs a eau bouillante

Energy Technology Data Exchange (ETDEWEB)

Allais, V

1998-03-01

This thesis was carried out in the framework of fuel cycles studies in partnership with COGEMA; the aim is to determine physics parameters characterising Boiling Reactor Assemblies. Those reactors Firstly distinguish themselves from Pressurised Water Reactor by the boiling of the moderator in the core and secondary by the strong neutronics heterogeneity due to complex design. The diphasic mixture formed is characterised by the void fraction parameter. The loss of information, and neutronic studies characteristics of Boiling Water Reactors led us to make preliminary studies having in view to quantify the void fraction impact on the isotopics evolution. Studies on neutronics influence of assemblies and control rods from the immediate environment allows to define the cluster size to describe. The radial description optimisation with APOLLO-2 is necessary to improve the calculation performance and to reduce the errors coming from the modelization. The following points were studied: pellet radial discretization, clustering of cells characterized by a similar behaviour, options in flux spatial calculation (interface current formalism), self-shielding optimisation (specific to each isotopes). The three dimensional modelization with CRONOS-2 and the simplified accounting of the thermohydraulics / neutronics coupling done by a procedure developed and written during this thesis, allow an evaluation of axial distribution of void fraction, power and burn-up during the irradiation. The comparison with experimental analytic results of complete assembly and pin samples dissolutions allows the qualification of this procedure and confirms the necessity to take into account the void fraction axial variation during the evolution. The application of an automatic coupling with the DARWIN cycle code will allow a precise burnup calculation to be utilized in an industrial procedure. (author)

Preparation fo nuclear research reactors operators

International Nuclear Information System (INIS)

Roedel, G.

1986-01-01

The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

Preparation of nuclear research reactors operators

International Nuclear Information System (INIS)

Roedel, G.

1986-01-01

The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN, are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

EDV supported dynamic fire protection concept adaptation during dismantling of nuclear facilities

International Nuclear Information System (INIS)

Mummert, Maxi; Traichel, Anke; Dilger, Matthias

2013-01-01

Fire protection concepts are supposed to be a decision guide for the definition of measures and priorities in fire fighting and fire prevention. In case of reactor dismantling a fire protection concept for the actual status is required. Following the fuel removal from the reactor the protection goals are reduced to the safe confinement of radioactive materials and the restriction of radiation exposure. A dynamic fire protection concept was developed to allow the compliance with the required protection measures with respect to the protection targets. The implementation of the dynamic fire protection concept simplifies the planning of the dismantling steps and to adjust the fire protection measured in the frame of changes in the plant.

Calculation of anti-seismic design for Xi'an pulsed reactor

International Nuclear Information System (INIS)

Li Shuian

2002-01-01

The author describes the reactor safety rule, safety regulation and design code that must be observed to anti-seismic design in Xi'an pulsed reactor. It includes the classification of reactor installation, determination of seismic loads, calculate contents, program, method, results and synthetically evaluation. According to the different anti-seismic structure character of reactor installation, an appropriate method was selected to calculate the seismic response. The results were evaluated synthetically using the design code and design requirement. The evaluate results showed that the anti-seismic design function of reactor installation of Xi'an pules reactor is well, and the structure integrality and normal property of reactor installation can be protect under the designed classification of the earthquake

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

RB research nuclear reactor, Annual report for 1981; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Sotic, O; Pesic, M; Vranic, S; Zivkovic, B; Bogdanovic, M; Petronijevic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1981-07-01

The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report.

Evaluation Of Fire Safety And Protection At PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Ahmad Nabil Ab Rahim; Alfred Sanggau Ligam; Nurhayati Ramli; Mohd Fazli Zakaria; Naim Syauqi Hamzah; Phongsakorn Prak; Mohammad Suhaimi Kassim; Zarina Masood

2014-01-01

Fire hazard is one of many risks that can affect the safety operation of PUSPATI TRIGA Reactor. Reactor building in Malaysian Nuclear Agency was built in 1980s and the fire system has been introduced since then. The evaluation of the fire safety system at this time is important to ensure the efficiency of fire prevention, fighting and mitigation task that probably occurs. This evaluation involves with the fire fighting system and equipment, integrity of the system from the perspective of management and equipment, fire fighting procedure and fire fighting response team. (author)

The nuclear reactor systems

International Nuclear Information System (INIS)

Bacher, P.

2008-01-01

This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

An Idea of 20% test of the Initial Core Reactor Physics

International Nuclear Information System (INIS)

Roh, Kyung Ho; Yang, Sung Tae; Jung, Ji Eun

2012-01-01

Many tests have been performed on the OPR1000 and APR1400 before commercial operation. Some of these tests were performed at reactor power levels of 20% and 50%. The CPC (Core Protection Calculator) power distribution test is one of these tests. It is performed to assure the reliability of the Core Protection Calculator System (CPCS). Through this test, SAM1 is calculated using the snapshots2. The test takes about nine hours at a reactor power level of 20% and about thirty hours at a reactor power level of 50%. SAM used at each reactor power level is as follows: 1. Reactor power of 0% ∼ 20%: Designed SAM (DSAM) 2. Reactor power of 20% ∼ 50%: SAM calculated (C-SAM) at a reactor power of 20% 3. Reactor power 50% ∼ End of Cycle : SAM calculated at a reactor power of 50% As mentioned earlier, SAM is calculated and punched into CPC to assure the reliability of CPCS. Therefore, CPC is operated having penalties with D-SAM until3 reaching a reactor power of 20%. That is, the penalty of CPC will be removed when SAM is calculated and punched into the CPC at a reactor power of 20%. But these penalties are considered to be removed after a reactor power of 50% test in order to maintain the conservatism of the CPC. This is done because the final values calculated using C-SAM, in contrast to those calculated using SAM, a reactor power of 50%, are not correct. This paper began from an idea, 'If so, what would happen if we removed the CPC power distribution test at a reactor power of 20%?'

Heavy water reactors physics; Physique des reacteurs a eau lourde

Energy Technology Data Exchange (ETDEWEB)

Girard, Y; Lourme, P; Naudet, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors) [French] Un important programme d'etudes sur la physique des reacteurs a eau lourde est mene en France depuis assez longtemps. La decision de construire le prototype EL 4 et de s'engager ainsi dans la filiere des reacteurs a eau lourde refroidis par gaz a redonne un nouvel interet a ce programme et l'a en meme temps oriente dans une direction plus particuliere. La presente communication, rassemble les resultats des etudes faites dans ce domaine depuis la derniere conference de Geneve. Dans la premiere partie on decrit les etudes experimentales dont la plupart ont ete effectuees dans la pile d'experiences critiques Aquilon II. Les experiences sont groupees en quatre ensembles: etude systematique de reseaux (mesures de laplaciens) etudes

Standard Technical Specifications for Westinghouse pressurized water reactors

International Nuclear Information System (INIS)

Virgilio, M.J.

1980-09-01

The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

Nuclear reactor with several cores

International Nuclear Information System (INIS)

Swars, H.

1977-01-01

Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

Backfitting of the FRG reactors

Energy Technology Data Exchange (ETDEWEB)

Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

1990-05-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

Backfitting of the FRG reactors

International Nuclear Information System (INIS)

Krull, W.

1990-01-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

Conceptual designs of power tokamak-type thermonuclear reactors

International Nuclear Information System (INIS)

Shejndlin, A.E.; Nedospasov, A.V.

1978-01-01

Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

Design of radiation shields in nuclear reactor core

International Nuclear Information System (INIS)

Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

2008-01-01

This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

Radiation protection/shield design: a need for a systems approach

International Nuclear Information System (INIS)

Disney, R.K.

1977-01-01

Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. The system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection is described, and the program developed to implement this approach is defined. In addition, the principal shielding design problems for LMFBR nuclear reactor systems are discussed in relation to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods are discussed

Comparison of performance indicators of different types of reactors based on ISOE database

International Nuclear Information System (INIS)

Janzekovic, H.; Krizman, M.

2005-01-01

The optimisation of the operation of a nuclear power plant (NPP) is a challenging issue due to the fact that besides general management issues, a risk associated to nuclear facilities should be included. In order to optimise the radiation protection programmes in around 440 reactors in operation with more than 500 000 monitored workers each year, the international exchange of performance indicators (PI) related to radiation protection issues seems to be essential. Those indicators are a function of a type of a reactor as well as the age and the quality of the management of the reactor. in general three main types of radiation protection PI could be recognised. These are: occupational exposure of workers, public exposure and management of PI related to radioactive waste. The occupational exposure could be efficiently studied using ISOC database. The dependence of occupational exposure on different types of reactors, e.g. PWR, BWR, are given, analysed and compared. (authors)

Primary system thermal-hydraulic simulation of a experimental pool type research fast reactor

International Nuclear Information System (INIS)

Borges, E.M.; Braz Filho, F.A.

1993-01-01

The first step of the Fast Reactor Program (REARA) is the design of an experimental reactor. To this end a 5 MW t pool type reactor was adapted. The objective of this work is to evaluate the reactor behaviour at the on set protected accidents. The program NALAP was used in this study and the results showed the outstanding safety margins that this reactor type presents inherently. (author)

MEANS FOR SHIELDING AND COOLING REACTORS

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Aspects of 238Pu production in the experimental fast reactor JOYO

International Nuclear Information System (INIS)

Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

2005-01-01

Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

Software verification and validation methodology for advanced digital reactor protection system using diverse dual processors to prevent common mode failure

International Nuclear Information System (INIS)

Son, Ki Chang; Shin, Hyun Kook; Lee, Nam Hoon; Baek, Seung Min; Kim, Hang Bae

2001-01-01

The Advanced Digital Reactor Protection System (ADRPS) with diverse dual processors is being developed by the National Research Lab of KOPEC for ADRPS development. One of the ADRPS goals is to develop digital Plant Protection System (PPS) free of Common Mode Failure (CMF). To prevent CMF, the principle of diversity is applied to both hardware design and software design. For the hardware diversity, two different types of CPUs are used for Bistable Processor and Local Coincidence Logic Processor. The VME based Single Board Computers (SBC) are used for the CPU hardware platforms. The QNX Operating System (OS) and the VxWorks OS are used for software diversity. Rigorous Software Verification and Validation (V and V) is also required to prevent CMF. In this paper, software V and V methodology for the ADRPS is described to enhance the ADRPS software reliability and to assure high quality of the ADRPS software

Trans-Uranium Doping Utilization for Increasing Protected Plutonium Proliferation of Small Long Life Reactor

Energy Technology Data Exchange (ETDEWEB)

Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suud, Zaki [Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suzuki, Mitsutoshi [Japan Atomic Energy Agency, Nuclear Non-proliferation Science and Technology Center, 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan)

2009-06-15

Scientific approaches are performed by adopting some methodologies in order to increase a material 'barrier' in plutonium isotope composition by increasing the even mass number of plutonium isotope such as Pu-238, Pu-240 and Pu-242. Higher difficulties (barrier) or more complex requirement for peaceful use of nuclear materials, material fabrication and handling and isotopic enrichment can be achieved by a higher isotopic barrier. Higher barrier which related to intrinsic properties of plutonium isotopes with even mass number (Pu-238, Pu-240 and Pu-242), in regard to their intense decay heat (DH) and high spontaneous fission neutron (SFN) rates were used as a parameter for improving the proliferation resistance of plutonium itself. Pu-238 has relatively high intrinsic characteristics of DH (567 W/kg) and SFN rate of 2660 n/g/s can be used for making a plutonium characteristics analysis. Similar characteristics with Pu-238, other even mass number of plutonium isotopes such as Pu-240 and Pu-242 have been shown in regard to SFN values. Those even number mass of plutonium isotope contribute to some criteria of plutonium characterization which will be adopted for present study such as IAEA, Pellaud and Kessler criteria (IAEA, 1972; Pellaud, 2002; and Kessler, 2004). The study intends to evaluate the trans-uranium doping effect for increasing protected plutonium proliferation in long-life small reactors. The development of small and medium reactor (SMR) is one of the options which have been adopted by IAEA as future utilization of nuclear energy especially for less developed countries (Kuznetsov, 2008). The preferable feature for small reactors (SMR) is long life operation time without on-site refueling and in the same time, it includes high proliferation resistance feature. The reactor uses MOX fuel as driver fuel for two different core types (inner and outer core) with blanket fuel arrangement. Several trans-uranium doping and some doping rates are evaluated

Experiments on light water lattices with enriched uranium fuel; Analyse des donnees experimentales sur les reseaux a eau legere et uranium enrichi

Energy Technology Data Exchange (ETDEWEB)

Audinet, M [Societe des Forges et Ateliers du Creusot, 75 - Paris (France); Lamare, J de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Panossian, J [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

Experiments a light water lattices with slightly enriched uranium fuel, have been performed at Brookhaven and Bettis Plant Laboratories. The results are studied and compared with simple theories on reactor calculations. By taking into account shadow effects and non Maxwellian neutron spectrum, which are important in this kind of reactors, we have been able to explain the observed results fairly well. We can thus give a constituent set of formulas with which to calculate lattices similar to there we studied. (author) [French] Les resultats d'experiences effectuees aux Laboratoires de Brookbaven et de Bettis Plant, sur des reseaux heterogenes a eau legere et uranium metallique legerement enrichi, sont analyses et confrontes avec les theories simples du calcul de pile. En tenant compte des effets d'interaction et d'echauffement du spectre de neutrons qui sont importants dans ce type de reacteurs, on parvient a rendre compte convenablement des resultats observes. On a ainsi mis au point un formulaire permettant le calcul des reseaux quivpeuvent etre consideres comme assez semblables aux reseaux etudies. (auteur)

MAPLE-X10 reactor digital control system

International Nuclear Information System (INIS)

Deverno, M.T.; Hinds, H.W.

1991-10-01

The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MW t , pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability

Long-lived Hybrid Incore Detector for Core Monitoring and Protection

Energy Technology Data Exchange (ETDEWEB)

Cha, Kyoon Ho [KEPCO Research Institute, Daejeon (Korea, Republic of)

2010-10-15

The signal production mechanism in a rhodium (Rh) fixed in-core detector emitter relies primarily on the beta particles resulting from neutron absorptions in either of two Rh isotopes to produce an electric current. As the neutron transmutation process depletes the Rh isotopes, the signal output per unit neutron flux from an Rh detector emitter will decrease. A vanadium detector is primarily sensitive to neutrons, but with a somewhat slower reaction time as that of a Rh detector. The benefit of vanadium over rhodium is its low depletion rate, which is a factor of 7 times less than that of rhodium. Platinum detectors are very sensitive to gamma flux, but only mildly sensitive to neutron flux. Because the depletion rate of platinum is very small, it can be neglected. Generally, both gamma and neutron signals are proportional to the assembly power. The characteristics of a new detector are the long life time due to the low depletion of emitter materials and the capability of reactor protection as well as reactor monitoring. The new detector uses vanadium and platinum as the emitter materials to meet the long life time and reactor protection capability. Vanadium detector is used for reactor monitoring and platinum detector is used for reactor protection. To determine the number of emitter strings, a comparative study of the power peaking factor monitoring accuracy for various self-powered fixed in-core detector geometries was made, and the configuration of the optimal detector design was also established and verified. The design of a new detector consists of five-string vanadium detector elements, and three-string platinum detector elements. The detector assembly also contains a background wire for compensation of noise signal and a thermocouple for use in the post-accident monitoring system. This new hybrid detector can be used for both reactor Monitoring And reactor Protection (MAP)

Aircraft-crash-protected steel reactor building roof structure for the European market

International Nuclear Information System (INIS)

Posta, B.A.; Kadar, I.; Rao, A.S.

1996-01-01

This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool

A new safety approach in the design of fast reactors

International Nuclear Information System (INIS)

Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

1987-01-01

A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

Reactor safeguards against insider sabotage

International Nuclear Information System (INIS)

Bennett, H.A.

1982-03-01

A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested

Method and device for the passive protection of a nuclear reactor

International Nuclear Information System (INIS)

Cachera, P.C.

1976-01-01

Conventional fuel elements within the core of a nuclear reactor and especially a fast reactor are at least partly replaced by ''safety elements'' each formed by a stack of fissile fuel pellets enclosed in a can. Each pellet is provided with a central orifice so as to form an axial flow duct of sufficiently large cross-sectional area to ensure that the portion of fuel which is liable to melt as a result of a neutron-flux excursion flows under gravity to the bottom of the fuel element and has the effect of reducing the reactivity without damaging the fuel can

Problems of control of WWER-type pressurized water reactors (PWR's)

International Nuclear Information System (INIS)

Drab, F.; Grof, V.

1978-01-01

The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

Reactor modification, preparation and operation

International Nuclear Information System (INIS)

Weill, J.; Furet, J.; Baillet, J.; Donvez, G.; Duchene, J.; Gras, R.; Mercier, R.; Chenouard, J.; Leconte, J.

1962-01-01

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system

Reactor modification, preparation and operation

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

1962-03-01

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

Reactor modification, preparation and operation

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

1962-03-15

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

The present status and the prospect of China research reactors

International Nuclear Information System (INIS)

Yongmao, Z.; Yizheng, C.

1990-01-01

A total of 100 reactor operation years' experience of research reactors has now been obtained in China. The type and principal parameters of China research reactors and their operating status are briefly introduced in this paper. Chinese research reactors have been playing an important role in nuclear power and nuclear weapon development, industrial and agricultural production, medicine, basic and applied science research and environmental protection, etc. The utilization scale, benefits and achievements will be given. There is a good safety record in the operation of these reactors. A general safety review is discussed. The important incidents and accidents happening during a hundred reactor operating years are described and analyzed. China has the capability of developing any type of research reactor. The prospective projects are briefly introduced

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)

Application of Emergency Action Levels from Potential Release at Research Reactor HANARO

International Nuclear Information System (INIS)

Kim, Jongsoo; Lee, Goan Yub; Lee, Hae Choi; Kim, Bong Suk

2014-01-01

Execution of the protective action promptly is possible that Emergency Action Levels (EALs) must be established for a radiological release from nuclear facility. The EALs for electric power reactor are already developed and applied to recognize an emergency situation rapidly. Recently the IAEA published the safety report including the EALs for research reactor. This paper describes the EALs to apply for a potential release pathway at the research reactor HANARO. The results of table 1 and 2 will be higher than actual because the weather condition in real situation is difference. However, the EALs applying the potential stack release, ground release and site can be useful for research reactor HANARO making the emergency declaration. The EALs at the site boundary of the table 3 can be applied to protect the off-site public

New flux detectors for CANDU 6 reactors

International Nuclear Information System (INIS)

Cuttler, J.M.; Medak, N.

1992-06-01

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

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

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

1977-01-01

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

The French-German common safety approach for future reactors

International Nuclear Information System (INIS)

Birkhofer, A.; Chevet, P.F.

1995-01-01

A common safety approach has been defined for future electronuclear plants in the framework of the French-German European Pressurised water Reactor (EPR) project. Improvements in the domain of containment are required in future reactors conception to prevent any risk of core fusion under high and low pressure. Another objective is to reduce significantly the radioactive releases due to other accidents in order to reduce spatial and temporal environmental and human protection procedures. Protection against external aggressions (plane fall, explosions, earthquakes,..), prevention of pipe rupture in the primary circuits, limitation of hydrogen production in the case of water-zirconium complete reaction, cooling of the reactor in the case of core fusion, and radiologic consequences of accidents are the main points discussed by the French-German safety authorities to define the common safety standards of the EPR project. (J.S.)

Nuclear fast neutron reactor cooled by a liquid metal and of which internal structures are equipped with a thermal protection device

International Nuclear Information System (INIS)

Lemercier, G.; Lions, N.

1986-01-01

The internal structures of a nuclear fast neutron reactor are covered at least partially, on the most hot side, by a thermal protection device. This device comprises modular plates arranged end to end with a certain play between themselves and taking approximately the shape of the internal structures. Each plate is fixed in its center on the internal structures by a stud. A small plate fixed at one of the corners of each plate and covering partially the adjacent plates ensures the safety fixing of these ones [fr

SUPERPHENIX: Reactor core temperatures survey by minicomputers - original aspects related to safety

International Nuclear Information System (INIS)

Berlin, C.; Josue, M.; Pinoteau, J.

1986-01-01

The system for core temperatures fast processing (TRIC) utilized in SUPERPHENIX is part of the reactor protection system. Due to the number of temperature measurements taken into account, to the specific data processing and to the rapidity required in the treatment, the use of digital computing devices is justified. The present paper describes the conception of the system in order to satisfy the special requirements for the computers used in power reactors protection systems

Dosimetry and technical radiation protection at the RA reactor in 1979, Report on the project 'Operation and maintenance of the RA reactor'

International Nuclear Information System (INIS)

Ninkovic, M. et al.

1979-12-01

This report include the analysis of the dosimetry and technical radiation protection results collected during 1979. The first part shows the data about the fundamental exposure to radiation and the statistical review of the total number of measurements. The following are included as well: measured values of radioactive gases and aerosol contents in air; contamination level of surfaces, clothes and uncovered parts of the staff bodies. Analysis of the personnel exposure is presented in the second part of the report. It was stated that the maximum individual external dose was 9.5 mGy, and the exposure dose was 1/5 lower than the annual dose limit for all the exposed personnel. It was estimated that that the internal contamination is negligible compared to the external contamination. This statement was based on the frequency of tasks in the contaminated zones, undertaken control and protection measures, as well as on the evaluation of the expected internal contamination. Comparative evaluation of the occupational exposures is given for the past five years. This showed that the exposure in 1979 was approximately on the same level as the mean value during past four years. The third part of the report covers the numerical data about the quantity of the collected radioactive waste, total size of contaminated and decontaminated surfaces and number of decontaminated objects. A brief analysis of the accidents occurred during this year under regular reactor operating conditions, maintenance and repair is given at the end. It was stated that there has been no accident that would cause exposure higher than the prescribed limits or any significant contamination of the working space or environment. The more detailed analyses concerned with radiation protection aspects of the irregularities noticed on the nuclear fuel elements were described in the Annex. It was concluded that the mentioned problems did not cause any leaks. This was significant for planning and manipulating the

Formal verification and validation of the safety-critical software in a digital reactor protection system

International Nuclear Information System (INIS)

Kwon, K. C.; Park, G. Y.

2006-01-01

This paper describes the Verification and Validation (V and V) activities for the safety-critical software in a Digital Reactor Protection System (DRPS) that is being developed through the Korea nuclear instrumentation and control system project. The main activities of the DRPS V and V process are a preparation of the software planning documentation, a verification of the software according to the software life cycle, a software safety analysis and a software configuration management. The verification works for the Software Requirement Specification (SRS) of the DRPS consist of a technical evaluation, a licensing suitability evaluation, a inspection and traceability analysis, a formal verification, and preparing a test plan and procedure. Especially, the SRS is specified by the formal specification method in the development phase, and the formal SRS is verified by a formal verification method. Through these activities, we believe we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the nuclear safety-critical software in a DRPS. (authors)

Software Unit Testing during the Development of Digital Reactor Protection System of HTR-PM

International Nuclear Information System (INIS)

Guo Chao; Xiong Huasheng; Li Duo; Zhou Shuqiao; Li Jianghai

2014-01-01

Reactor Protection System (RPS) of High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) is the first digital RPS designed and to be operated in the Nuclear Power Plant (NPP) of China, and its development process has receives a lot of concerns around the world. As a 1E-level safety system, the RPS has to be designed and developed following a series of nuclear laws and technical disciplines including software verification and validation (software V&V). Software V&V process demonstrates whether all stages during the software development are performed correctly, completely, accurately, and consistently, and the results of each stage are testable. Software testing is one of the most significant and time-consuming effort during software V&V. In this paper, we give a comprehensive introduction to the software unit testing during the development of RPS in HTR-PM. We first introduce the objective of the testing for our project in the aspects of static testing, black-box testing, and white-box testing. Then the testing techniques, including static testing and dynamic testing, are explained, and the testing strategy we employed is also introduced. We then introduce the principles of three kinds of coverage criteria we used including statement coverage, branch coverage, and the modified condition/decision coverage. As a 1E-level safety software, testing coverage needs to be up to 100% mandatorily. Then we talk the details of safety software testing during software development in HTR-PM, including the organization, methods and tools, testing stages, and testing report. The test result and experiences are shared and finally we draw a conclusion for the unit testing process. The introduction of this paper can contribute to improve the process of unit testing and software development for other digital instrumentation and control systems in NPPs. (author)

Next generation advanced nuclear reactor designs

International Nuclear Information System (INIS)

Turgut, M. H.

2009-01-01

Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

Radiation protection programme for a radioisotope production facility

International Nuclear Information System (INIS)

Makgato, Thutu Nelson

2015-02-01

The present project reviews reactor based radioisotope production facilities. An overview of techniques and methodologies used as well as laboratory facilities necessary for the production process are discussed. Specific details of reactor based production and processing of more commonly used industrial and pharmaceutical radioisotopes are provided. Ultimately, based on facilities and techniques utilized as well as the associated hazard assessment, a proposed radiation protection programme is discussed. Elements of the radiation protection programme will also consider lessons from recent incidents and accidents encountered in radioisotope production facilities. (au)

Fluidized bed nuclear reactor as a IV generation reactor

International Nuclear Information System (INIS)

Sefidvash, Farhang

2002-01-01

The object of this paper is to analyze the characteristics of the Fluidized Bed Nuclear Reactor (FBNR) concept under the light of the requirements set for the IV generation nuclear reactors. It is seen that FBNR generally meets the goals of providing sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production; minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for the public health and the environment; increase the assurance that it is a very unattractive and least desirable route for diversion or theft of weapons-usable materials; excel in safety and reliability; have a very low likelihood and degree of reactor core damage; eliminate the need for offsite emergency response; have a clear life-cycle cost advantage over other energy sources; have a level of financial risk comparable to other energy projects. The other advantages of the proposed design are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. (author)

Tritium protective clothing

International Nuclear Information System (INIS)

Fuller, T.P.; Easterly, C.E.

1979-06-01

Occupational exposures to radiation from tritium received at present nuclear facilities and potential exposures at future fusion reactor facilities demonstrate the need for improved protective clothing. Important areas relating to increased protection factors of tritium protective ventilation suits are discussed. These areas include permeation processes of tritium through materials, various tests of film permeability, selection and availability of suit materials, suit designs, and administrative procedures. The phenomenological nature of film permeability calls for more standardized and universal test methods, which would increase the amount of directly useful information on impermeable materials. Improvements in suit designs could be expedited and better communicated to the health physics community by centralizing devlopmental equipment, manpower, and expertise in the field of tritium protection to one or two authoritative institutions

Tritium protective clothing

Energy Technology Data Exchange (ETDEWEB)

Fuller, T. P.; Easterly, C. E.

1979-06-01

Occupational exposures to radiation from tritium received at present nuclear facilities and potential exposures at future fusion reactor facilities demonstrate the need for improved protective clothing. Important areas relating to increased protection factors of tritium protective ventilation suits are discussed. These areas include permeation processes of tritium through materials, various tests of film permeability, selection and availability of suit materials, suit designs, and administrative procedures. The phenomenological nature of film permeability calls for more standardized and universal test methods, which would increase the amount of directly useful information on impermeable materials. Improvements in suit designs could be expedited and better communicated to the health physics community by centralizing devlopmental equipment, manpower, and expertise in the field of tritium protection to one or two authoritative institutions.

Development of physical conceptions of fast reactors

International Nuclear Information System (INIS)

Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

2013-01-01

• Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

1997-01-01

In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

Advanced power reactors with improved safety characteristics

International Nuclear Information System (INIS)

Birkhofer, A.

1994-01-01

The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

Radiological protection of the staff during the decommissioning operations of the Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Ene, D.C.

2002-01-01

Dose rate estimates for periods of 100 days and 6, 10, 25, 100 years after the shut down of the Romanian VVR-S reactor are presented in this paper for some foreseen decommissioning activities which include: i) cutting the water pipe in the pump room and the reactor sealing operations; ii) extracting reactor components; and iii) handling and dismantling the internal structures taken of from the reactor. For the reactor components extracted from the reactor, the considered calculation points were placed in the central plan of the items, on the surface and at distances from the surface which correspond to +0.2m, +1m, +2m, +8m, and +10m. Time dependence of the resulted dose rates are presented and discussed. Qualitative comparison with the measured values from other VVR-S reactors is done. The obtained results assist to develop working procedures that must be observed during the decommissioning activities. (author)

Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

International Nuclear Information System (INIS)

Vito, D.J.

1980-12-01

The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

The improvement of the fire protections system for nuclear cycle facilities. Formulation of a fire protection guideline for nuclear fuel cycle facilities

International Nuclear Information System (INIS)

2012-04-01

The private side Fire Protection Guideline was investigated with respect to the fire having taken place at the nuclear reactor site followed by the Chuetsu-Oki earthquake in Niigata Prefecture in 2007. To improve the fire protection system especially applicable to MOX fuel fabrication facilities, JNES (Japan Nuclear Energy Safety Organization) investigated private guidelines adopted in Japanese Light Water cooled Reactors, the standardized guidelines used in Nuclear Facilities in other countries including USA, and the standards in the chemical plants. The content of the guideline concerns the prevention of the fire breakout, the prevention of fire extension, the reduction of the fire effects, as well as the facility-characteristic protection countermeasures and the fire effect evaluations. (S. Ohno)

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1962-09-15

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined.

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

Energy Technology Data Exchange (ETDEWEB)

Moses, David Lewis [ORNL

2011-10-01

This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

A Compact Quasi-axisymmetric Stellarator Reactor

International Nuclear Information System (INIS)

Ku, L.P.

2003-01-01

We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils

Decommissioning a nuclear reactor

International Nuclear Information System (INIS)

Montoya, G.M.

1991-01-01

The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs