Application of Pressure Equipment Standard at nuclear power plants; Aplicacion del Reglamento de Equipos a Presion a las centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Mostaza, J. M.

2011-07-01

Regarding with the paper presented on 9{sup t}h June 2011 referred to the Industrial Security standard in Nuclear Plants, it was about the application of Pressure Equipment standard to mentioned Nuclear Plants, this article is an extract of the paper going to be exposed. (Author)

Application on electrochemistry measurement of high temperature high pressure condition in PWR nuclear power plants

International Nuclear Information System (INIS)

Li Yuchun; Xiao Zhongliang; Jiang Ya; Yu Xiaowei; Pang Feifei; Deng Fenfang; Gao Fan; Zhou Nianguang

2011-01-01

High temperature high pressure electrochemistry testing system was comprehensively analyzed in this paper, according to actual status for supervision in primary and secondary circuits of PWR nuclear power plants. Three research methods were reviewed and discussed for in-situ monitor system. By combination with ECP realtime measurement it was executed for evaluation and water chemistry optimization in nuclear power plants. It is pointed out that in-situ electrochemistry measurement has great potential application for water chemistry evaluation in PWR nuclear power plants. (authors)

Modeling and simulation of pressurizer dynamic process in PWR nuclear power plant

International Nuclear Information System (INIS)

Ma Jin; Liu Changliang; Li Shu'na

2010-01-01

By analysis of the actual operating characteristics of pressurizer in pressurized water reactor (PWR) nuclear power plant and based on some reasonable simplification and basic assumptions, the quality and energy conservation equations about pressurizer' s steam zone and the liquid zone are set up. The purpose of this paper is to build a pressurizer model of two imbalance districts. Water level and pressure control system of pressurizer is formed though model encapsulation. Dynamic simulation curves of main parameters are also shown. At last, comparisons between the theoretical analysis and simulation results show that the pressurizer model of two imbalance districts is reasonable. (authors)

General requirements for pressure-retaining systems and components in CANDU nuclear power plants

International Nuclear Information System (INIS)

1991-11-01

This standard specifies the general requirements for the design, fabrication and installation of pressure-retaining systems, components, and their supports in CANDU nuclear power plants. (16 figs., 2 tabs., 25 refs.)

On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

International Nuclear Information System (INIS)

Pan Xiren; Zhang Chen.

1988-01-01

The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

On-line testing of nuclear plant temperature and pressure instrumentation and other critical plant equipment. IAEA regional workshop. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-12-31

Under European regional TC project RER/4/011, IAEA and VUJE Training centre organized a workshop on On-line Testing of Nuclear Power Plant Temperature and Pressure Instrumentation and Other Critical Plant Equipment in Trnava, Slovak Republic, from 25 to 29 May 1998. The objective of the workshop was to review the state-of-the-art in NPP instrumentation, cover typical instrumentation problems and solutions, describe technical and regulatory requirements for verifying the performance of nuclear power plant instrumentation, describe new methods developed and applied in NPPs for on-line verification and performance of instrumentation and present new techniques using existing instrumentation to identify the on-set problems in the plant electrical, mechanical and thermal hydraulic systems. Particular emphasis was placed on temperature measurements by Resistance Temperature Detectors (RTDs) and thermocouples and pressure measurements using motion-balanced and forced-balanced pressure transmitters. This proceedings includes papers presented by the invited speakers and the participants each with an abstract as wells as a summary of the Round-Table discussions Refs, figs, tabs

On-line testing of nuclear plant temperature and pressure instrumentation and other critical plant equipment. IAEA regional workshop. Working material

International Nuclear Information System (INIS)

1998-01-01

Under European regional TC project RER/4/011, IAEA and VUJE Training centre organized a workshop on On-line Testing of Nuclear Power Plant Temperature and Pressure Instrumentation and Other Critical Plant Equipment in Trnava, Slovak Republic, from 25 to 29 May 1998. The objective of the workshop was to review the state-of-the-art in NPP instrumentation, cover typical instrumentation problems and solutions, describe technical and regulatory requirements for verifying the performance of nuclear power plant instrumentation, describe new methods developed and applied in NPPs for on-line verification and performance of instrumentation and present new techniques using existing instrumentation to identify the on-set problems in the plant electrical, mechanical and thermal hydraulic systems. Particular emphasis was placed on temperature measurements by Resistance Temperature Detectors (RTDs) and thermocouples and pressure measurements using motion-balanced and forced-balanced pressure transmitters. This proceedings includes papers presented by the invited speakers and the participants each with an abstract as wells as a summary of the Round-Table discussions

Integrated Nuclear Recycle Plant

International Nuclear Information System (INIS)

Patodi, Anuj; Parashar, Abhishek; Samadhiya, Akshay K.; Ray, Saheli; Dey, Mitun; Singh, K.K.

2017-01-01

Nuclear Recycle Board (NRB), Tarapur proposes to set up an 'Integrated Nuclear Recycle Plant' at Tarapur. This will be located in the premises of BARC facilities. The project location is at coastal town of Tarapur, 130 Km north of Mumbai. Project area cover of INRP is around 80 hectares. The plant will be designed to process spent fuel received from Pressurized Heavy Water Reactors (PHWRs). This is the first large scale integrated plant of the country. INRP will process spent fuel obtained from indigenous nuclear power plants and perform left over nuclear waste disposal

Effects of normal aging on calibration and response time of nuclear plant RTDs and pressure sensors

International Nuclear Information System (INIS)

Hashemian, H.L.; Riner, J.L.

1993-01-01

Resistance temperature detectors (RTDs) and pressure, level, and flow transmitters provide a majority of the vital signals for the control and safety of nuclear power plants. Therefore, it is crucial to ensure that the performance of these sensors are maintained at an acceptable level while the plant is operating. Since aging has the potential to cause performance degradation in RTDs and pressure transmitters, several research projects have been sponsored by the US Nuclear Regulatory Commission (NRC) to study the aging characteristics of these sensors and ensure that adequate test methods and test frequencies are followed by the nuclear industry to ensure safety. The details of these projects are summarized in this paper

Nuclear power plants

International Nuclear Information System (INIS)

Margulova, T.Ch.

1976-01-01

The textbook focuses on the technology and the operating characteristics of nuclear power plants equiped with pressurized water or boiling water reactors, which are in operation all over the world at present. The following topics are dealt with in relation to the complete plant and to economics: distribution and consumption of electric and thermal energy, types and equipment of nuclear power plants, chemical processes and material balance, economical characteristics concerning heat and energy, regenerative preheating of feed water, degassing and condenser systems, water supply, evaporators, district heating systems, steam generating systems and turbines, coolant loops and pipes, plant siting, ventilation and decontamination systems, reactor operation and management, heat transfer including its calculation, design of reactor buildings, and nuclear power plants with gas or sodium cooled reactors. Numerous technical data of modern Soviet nuclear power plants are included. The book is of interest to graduate and post-graduate students in the field of nuclear engineering as well as to nuclear engineers

Pressurized thermal shock analysis in German nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fricke, Stefan; Braun, Michael [TUEV NORD Nuclear, Hannover (Germany)

2015-03-15

For more than 30 years TUeV NORD is a competent consultant in nuclear safety is-sues giving expert third party opinion to our clients. According to the German regulations the safety against brittle fracture has to be proved for the reactor pressure vessel (RPV) and with a new level of knowledge the proof has to be continuously updated with the development in international codes and standards like ASME, BS and RCC-M. The load of the RPV is a very complex transient pressure and temperature situation. Today these loading conditions can be modeled by thermal hydraulic calculations and new experimental results much more detailed than in the construction phase of German Nuclear Power Plants in the 1980s. Therefore, the proof against brittle fracture from the construction phase had to be updated for all German Nuclear Power Plants with the new findings of the loading conditions especially for a postulated small leakage in the main coolant line. The RPV consists of ferritic base material (about 250 mm) and austenitic cladding (about 6 mm) at the inner side. The base material and the cladding have different physical properties which have to be considered temperature dependently in the cal-culations. Radiation-embrittlement effects on the material are to be respected in the fracture mechanics assessment. The regions of the RPV of special interest are the core weld, the inlet and outlet nozzle region and the flange connecting weld zone. The fracture mechanics assessment is performed for normal and abnormal operating conditions and for accidents like LOCA (Loss of Coolant Accident). In this paper the German approach to fracture mechanics assessment to brittle fracture will be discussed from the point of view of a third party organization.

A survey of repair practices for nuclear power plant containment metallic pressure boundaries

Energy Technology Data Exchange (ETDEWEB)

Oland, C.B.; Naus, D.J. [Oak Ridge National Lab., TN (United States)

1998-05-01

The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition.

A survey of repair practices for nuclear power plant containment metallic pressure boundaries

International Nuclear Information System (INIS)

Oland, C.B.; Naus, D.J.

1998-05-01

The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition

Excessive leakage measurement using pressure decay method in containment building local leakage rate test at nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Kyu; Kim, Chang Soo; Kim, Wang Bae [KHNP, Central Research Institute, Daejeon (Korea, Republic of)

2016-06-15

There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

Nuclear power plant

International Nuclear Information System (INIS)

Aisaka, Tatsuyoshi; Kamahara, Hisato; Yanagisawa, Ko.

1982-01-01

Purpose: To prevent corrosion stress cracks in structural materials in a BWR type nuclear power plant by decreasing the oxygen concentration in the reactor coolants. Constitution: A hydrogen injector is connected between the condensator and a condensate clean up system of a nuclear power plant. The injector is incorporated with hydrogenated compounds formed from metal hydrides, for example, of alloys such as lanthanum-nickel alloy, iron titanium alloy, vanadium, palladium, magnesium-copper alloy, magnesium-nickel alloy and the like. Even if the pressure of hydrogen obtained from a hydrogen bomb or by way of water electrolysis is changed, the hydrogen can always be injected into a reactor coolant at a pressure equal to the equilibrium dissociation pressure for metal hydride by introducing the hydrogen into the hydrogen injector. (Seki, T.)

A technical learning on the Pressurized Water Nuclear Power Plants using animation

International Nuclear Information System (INIS)

Ito, Hajime; Tomohara, Yasutaka; Kubo, Setsuo; Ninomiya, Toshiaki

2002-01-01

The pressurized water nuclear power generation plants tends to reduce construction of its new plant from viewpoints of recent stabilization in power demand/supply balance, development of new siting points, and so on. And, together with reducing any opportunity to experience at site, generation alternation to younger engineers without such experiences is progressing. In order to carry out technical tradition with high quality , as it is important to understand experiences of troubles and so on as valuable inheritance to apply them to actual use, it can be thought, in doubt, to be one of solving measures to prepare some learning tools applying the newest technology. The Kansai Electric Co., Ltd. Developed a CAD software using animation and 3D pictures using a personal computer which is edited some processes of technical transition on nuclear energy as a reference on a shape of CD ROM as an object from initial period of nuclear power station to present APWR. (G.K.)

Evaluation of a Kalman filter based power pressurizer instrument failure detection system implemented on a nuclear power plant training simulator

International Nuclear Information System (INIS)

Seegmiller, D.S.

1984-01-01

The usefulness of a nuclear power plant training simulator for developing and testing modern estimation and control applications for nuclear power plants is demonstrated. A Kalman filter based instrument failure detection technique for a pressurized water reactor pressurizer is implemented on the Department of Energy N Reactor Training Simulator. This real-time failure detection method computes the first two moments (mean and variance) of each element of a normalized filter innovations vector. Failed pressurizer instrumentation can be detected by comparing these moments to the known statistical properties of the steady state, linear Kalman fitler innovations sequence. The capabilities of the detection system are evaluated using simulated plant transients and instrument failures

Operation and maintenance of nuclear power plants

International Nuclear Information System (INIS)

Ackermann, G.

1987-01-01

This textbook gives a systematic introduction into the operational and maintenance activities in nuclear power plants with pressurized water reactors. Subjects: (1) Setup and operational behaviour of power reactors, (2) setup of nuclear power plants, (3) radiation protection and nuclear safety, (4) nuclear fuel, (5) constructional layout of nuclear power plants, (6) management, and (7) maintenance. 158 figs., 56 tabs

Effects of aging on calibration and response time of nuclear plant pressure transmitters

International Nuclear Information System (INIS)

Hashemian, H.M.

1991-01-01

This paper presents the key results of an experimental research project conducted for the Nuclear Regulatory Commission to quantify the effects of normal aging on static and dynamic performance of nuclear grade pressure, level, and flow transmitters (hereafter referred to as pressure transmitters). The project involved laboratory testing of representative pressure transmitters manufactured by Barton, Foxboro, Rosemount, and Tobar (or Veritrak) companies. These manufacturers provide the four most commonly used pressure transmitters in the safety systems of US nuclear power plants. The transmitters were tested under normal aging conditions as opposed to accelerated aging, even though accelerated aging will be used in the last few months of the project to determine the weak links and failure modes of the transmitters. The project has been performed in two phases. The Phase 1 project which was a six month feasibility study has been completed and the results published in NUREG/CR-5383. The Phase 2 project is still underway with the final report due in the fall of 1991. The project has focused on the following areas: (1) effects of aging on calibration stability; (2) effects of aging on response time; (3) study of individual components of pressure transmitters that are sensitive to aging degradation; (4) sensing line blockages due to solidification of boron, formation of sludge, freezing, and other effects; (5) search of licensee event reports and component reliability databases for failures of safety-related pressure transmitters; and (6) oil loss syndrome in Rosemount pressure transmitters

Role of non destructive techniques for monitoring structural integrity of primary circuit of pressurized water reactor nuclear power plant

International Nuclear Information System (INIS)

Sharma, P.K.; Sreenivas, P.

2015-01-01

The safety of nuclear installations is ensured by assessing status of primary equipment for performing the intended function reliably and maintaining the integrity of pressure boundaries. The pressure boundary materials undergo material degradation during the plant operation. Pressure boundary materials are subjected to operating stresses and material degradation that results in material properties changes, discontinuities initiation and increase in size of existing discontinuities. Pre-Service Inspection (PSI) is performed to generate reference base line data of initial condition of the pressure boundary. In-Service Inspections (ISI) are performed periodically to confirm integrity of pressure boundaries through comparison with respect to base line data. The non destructive techniques are deployed considering nature of the discontinuities expected to be generated through operating conditions and degradation mechanisms. The paper is prepared considering Pressurized Water Reactor (PWR) Nuclear Power Plant. The paper describes the degradation mechanisms observed in the PWR nuclear power plants and salient aspect of PSI and ISI and considerations in selecting non destructive testing. The paper also emphasises on application of acoustic emission (AE) based condition monitoring systems that can supplement in-service inspections for detecting and locating discontinuities in pressure boundaries. Criticality of flaws can be quantitatively evaluated by determining their size through in-service inspection. Challenges anticipated in deployment of AE based monitoring system and solutions to cater those challenges are also discussed. (author)

China’s Nuclear Power Plants in Operation

Institute of Scientific and Technical Information of China (English)

无

2011-01-01

Qinshan Plant Phase I Located in Haiyan,Zhejiang Province,Qinshan Nuclear Power Plant Phase I is t he first 300-megawatt pressurized water reactor (PWR) nuclear power plant independently designed,constructed,operated and managed by China.The plant came into commercial operation in April 1994.

Safety implications associated with in-plant pressurized gas storage and distribution systems in nuclear power plants

International Nuclear Information System (INIS)

Guymon, R.H.; Casto, W.R.; Compere, E.L.

1985-05-01

Storage and handling of compressed gases at nuclear power plants were studied to identify any potential safety hazards. Gases investigated were air, acetylene, carbon dioxide, chlorine, Halon, hydrogen, nitrogen, oxygen, propane, and sulfur hexaflouride. Physical properties of the gases were reviewed as were applicable industrial codes and standards. Incidents involving pressurized gases in general industry and in the nuclear industry were studied. In this report general hazards such as missiles from ruptures, rocketing of cylinders, pipe whipping, asphyxiation, and toxicity are discussed. Even though some serious injuries and deaths over the years have occurred in industries handling and using pressurized gases, the industrial codes, standards, practices, and procedures are very comprehensive. The most important safety consideration in handling gases is the serious enforcement of these well-known and established methods. Recommendations are made concerning compressed gas cylinder missiles, hydrogen line ruptures or leaks, and identification of lines and equipment

Reactor pressure vessel life cycle management at the Calvert Cliffs Nuclear Power Plant

International Nuclear Information System (INIS)

Doroshuk, B.W.; Bowman, M.E.; Henry, S.A.; Pavinich, W.A.; Lapides, M.E.

1993-01-01

Life Cycle Management (LCM) seeks to manage the aging process of important systems, structures, and components during licensed operation. The goal of Baltimore Gas and Electric Company's (BG and E) Life Cycle Management Program is to assure attainment of 40 years of operation and to preserve the option of an additional 20 years of operation for the Calvert Cliffs Nuclear Power Plant (CCNPP). Since the reactor pressure vessel (RPV) has been identified as one of the most critical components with regard to long-term operation of a nuclear power plant, BG and E initiated actions to manage life limiting or aging issues for the CCNPP RPVs. To achieve long-term operation, technical RPV issues must be effectively managed. This paper describes methods BG and E uses for managing RPV age-related degradation. (author)

Nuclear power plant pressurizer fault diagnosis using fuzzy signed-digraph and spurious faults elimination methods

International Nuclear Information System (INIS)

Park, Joo Hyun

1994-02-01

In this work, the Fuzzy Signed Digraph(FSD) method which has been researched for the fault diagnosis of industrial process plant systems is improved and applied to the fault diagnosis of the Kori-2 nuclear power plant pressurizer. A method for spurious faults elimination is also suggested and applied to the fault diagnosis. By using these methods, we could diagnose the multi-faults of the pressurizer and could also eliminate the spurious faults of the pressurizer caused by other subsystems. Besides the multi-fault diagnosis and system-wide diagnosis capabilities, the proposed method has many merits such as real-time diagnosis capability, independency of fault pattern, direct use of sensor values, and transparency of the fault propagation to the operators

Nuclear power plant pressurizer fault diagnosis using fuzzy signed-digraph and spurious faults elimination methods

International Nuclear Information System (INIS)

Park, Joo Hyun; Seong, Poong Hyun

1994-01-01

In this work, the Fuzzy Signed Digraph (FSD) method which has been researched for the fault diagnosis of industrial process plant systems is improved and applied to the fault diagnosis of the Kori-2 nuclear power plant pressurizer. A method for spurious faults elimination is also suggested and applied to the fault diagnosis. By using these methods, we could diagnose the multi-faults of the pressurizer and could also eliminate the spurious faults of the pressurizer caused by other subsystems. Besides the multi-fault diagnosis and system-wide diagnosis capabilities, the proposed method has many merits such as real-time diagnosis capability, independency of fault pattern, direct use of sensor values, and transparency of the fault propagation to the operators. (Author)

Nuclear Power Plant Lifetime Management Study (I)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

Nuclear Power Plant Lifetime Management Study (I)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-12-31

As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

Data list of nuclear power plants of pressurized-water reactor type in Japan

International Nuclear Information System (INIS)

Izumi, Fumio; Harayama, Yasuo

1981-08-01

This report has collected and compiled the data concerning performances, equipments and installations for nuclear power plants of the pressurized-water reactor type in Japan. The data used in the report are based on informations that were collected before December in 1980. The report is edited by modifing changes of the data appeared after publication of 1979 edition (JAERI-M 8947), and extending the data-package to cover new plants proposed thereafter. All data have been processed and tabulated with a computer program FREP, which has been developed as an exclusive use of data processing. (author)

Nuclear power plant diagnostic system

International Nuclear Information System (INIS)

Prokop, K.; Volavy, J.

1982-01-01

Basic information is presented on diagnostic systems used at nuclear power plants with PWR reactors. They include systems used at the Novovoronezh nuclear power plant in the USSR, at the Nord power plant in the GDR, the system developed at the Hungarian VEIKI institute, the system used at the V-1 nuclear power plant at Jaslovske Bohunice in Czechoslovakia and systems of the Rockwell International company used in US nuclear power plants. These diagnostic systems are basically founded on monitoring vibrations and noise, loose parts, pressure pulsations, neutron noise, coolant leaks and acoustic emissions. The Rockwell International system represents a complex unit whose advantage is the on-line evaluation of signals which gives certain instructions for the given situation directly to the operator. The other described systems process signals using similar methods. Digitized signals only serve off-line computer analyses. (Z.M.)

Nuclear power plant pressurizer fault diagnosis using fuzzy signed-digraph method

International Nuclear Information System (INIS)

Park, Joo Hyun; Seong, Poong Hyun

2004-01-01

In this study, The Fuzzy Signed Digraph method which has been researched and applied to the chemical process is improved and applied to the fault diagnosis of the pressurizer in nuclear power plants. The Fuzzy Signed-Digraph (FSD) is the method which applies the fuzzy number to the Signed-Digraph (SDG) method. The current SDG methods have many merits as follows: (1) SDG method can directly use the value of sensors not the alarm to the fault diagnosis. (2) This method can diagnose the fault independent on the pattern. (3) This method can diagnose the faults fastly because the method uses the cause-effect relation instead of the complex control equation among the variables. But, they are not proper to be applied to the diagnosis of the multi-faults and to diagnose faults on real time. It is because the unmeasured nodes in those methods must be connected to each other in order to find out the single fault under the single-fault assumption. These methods need long CPU time and cannot be applied to the multi-faults diagnosis. We propose a method in which the values of the unmeasured nodes are calculated from the relations between the unmeasured nodes and the measured nodes. By using this method, the CPU time for diagnosis can be reduced. This CPU time reduction makes the real-time diagnosis possible. This method can also be applied for the multi-faults diagnosis. This method is applied to the diagnosis of the pressurizer of the nuclear power plant KORI-2 in Korea. (author)

Underwater nuclear power plant structure

International Nuclear Information System (INIS)

Severs, S.; Toll, H.V.

1982-01-01

A structure for an underwater nuclear power generating plant comprising a triangular platform formed of tubular leg and truss members upon which are attached one or more large spherical pressure vessels and one or more small cylindrical auxiliary pressure vessels. (author)

Operation of Finnish nuclear power plants

International Nuclear Information System (INIS)

Tossavainen, K.

1992-03-01

The Finnish nuclear power plant units Loviisa 1 and 2 as well as TVO I and II were in operation for almost the whole third quarter of 1991. Longer interruptions in electricity generation were caused by the annual maintenances of the Loviisa plant units. The load factor average was 81.7 %. In a test conducted during the annual maintenance outage of Loviisa 1 it was detected that the check valve of the discharge line of one pressurized emergency make-up tank did not open sufficiently at the tank's hydrostatic pressure. In connection with a 1988 modification, a too tightly dimensioned bearing had been mounted on the valve's axle rod and the valve had not been duly tested after the operation. The event is classified as Level 1 on the International Nuclear Event Scale. Other events in this quarter which are classified according to the International Nuclear Event Scale are Level Zero (Below Scale). Occupational radiation doses and releases of radioactive material off-site were below authorised limits in this quarter. Only small amounts of radioactive materials originating in nuclear power plants were detected in samples taken in the vicinity of nuclear power plants

Partner of nuclear power plants

International Nuclear Information System (INIS)

Gribi, M.; Lauer, F.; Pauli, W.; Ruzek, W.

1992-01-01

Sulzer, the Swiss technology group, is a supplier of components and systems for nuclear power plants. Important parts of Swiss nuclear power stations, such as containments, reactor pressure vessels, primary pipings, are made in Winterthur. Sulzer Thermtec AG and some divisions of Sulzer Innotec focus their activities on servicing and backfitting nuclear power plants. The European market enjoys priority. New types of valves or systems are developed as economic solutions meeting more stringent criteria imposed by public authorities or arising from operating conditions. (orig.) [de

Finite element method used in strength calculations of nuclear power plant pressure vessels

International Nuclear Information System (INIS)

Hanulak, E.

1987-01-01

A software system based on the use of the finite element method in linear and nonlinear elastomechanics was developed for assessing the strength and service life of steam generators and pressurizers for WWER type nuclear power plants. The individual programs are briefly described. They are written in FORTRAN IV, some modules are in ASSEMBLER. Programs EGUSAP, NEANKO, ROSYNA are designed for the calculation of stress and deformation, programs ROSYNA, NEANKO and NTEPLO are used for the calculation of temperature fields. Programs SPOJ and STATES are used for assessing the strength and service life of screw joints and other nodes of the WWER-440 type steam generators and pressurizers. (Z.M.)

Multi-stage-flash desalination plants of relative small performance with integrated pressurized water reactors as a nuclear heat source

International Nuclear Information System (INIS)

Petersen, G.; Peltzer, M.

1977-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination plants with a performance in the range of 10 000 to 80 000 m 3 distillate per day heated by a nuclear reactor are investigated. The reactor concept is similar to the Integrated Pressurized Water Reactor (IPWR) of the nuclear ship OTTO HAHN. The design study shows that IPWR systems have specific advantages up to 200 MWth compared to other reactor types at least being adapted for single- and dual-purpose desalination plants. The calculated costs of the desalinated water show that due to fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (author)

The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

International Nuclear Information System (INIS)

Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

2012-01-01

This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

International Nuclear Information System (INIS)

2007-06-01

be used in reassessing the safety of individual operating plants. In 1998, the IAEA completed IAEA-TECDOC-1044 entitled Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for their Resolution and established the associated LWRGSIDB database (Computer Manual Series No. 13). The present compilation, which is based on broad international experience, is an extension of this work to cover pressurized heavy water reactors (PHWRs). As in the case of LWRs, it is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It addresses generic safety issues identified in nuclear power plants using PHWRs. In most cases, the measures taken or planned to resolve these issues are also identified. The work on this report was initiated by the Senior Regulators of Countries Operating CANDU-Type Nuclear Power Plants at one of their annual meetings. It was carried out within the framework of the IAEA's programme on National Regulatory Infrastructure for Nuclear Installation Safety and serves to enhance regulatory effectiveness through the exchange of safety related information

Commissioning of the nuclear power plant

International Nuclear Information System (INIS)

Furtado, P.M.; Rolf, F.

1984-01-01

Nuclear Power Plant Angra 2, located at Itaorna Beach-Angra dos Reis is the first plant of the Brazilian-German Agreement to be commissioned. The Nuclear Power Plant is a pressurized water reactor rated at 3765 Mw thermal/1325 Mw electrical. For commissioning purpose the plant is divided into 110 systems. Plant commissioning objective is to demonstrate the safe and correct operation of each plan component, system and of the whole plant in agreement with design conditions, licensing requirements and contractual obligations. This work gives a description of plant commissioning objectives, activities their time sequence, and documentation. (Author) [pt

A multi-stage-flash desalination plant of relative small performance with an integrated pressurized water reactor as a nuclear heat source

International Nuclear Information System (INIS)

Peltzer, M.; Petersen, G.

1976-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination-plants with a performance in the range of 10,000 to 80,000 m 3 /d heated by a nuclear reactor are investigated. The reactor concept is similar to the integrated pressurized water reactor (IPWR) of the nuclear ship OTTO HAHN. The calculated costs of the desalinated water show, that due to the fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (orig.) [de

Pressure suppression apparatus of a nuclear power plant

International Nuclear Information System (INIS)

Mizumachi, W.; Funalashi, T.

1980-01-01

Pressure suppression apparatus for a nuclear reactor comprises a vessel surrounding a reactor pressure vessel and containing a water pool at the bottom of the vessel, and a steam exhaust pipe. The apparatus further comprises an exhaust chamber connected to the immersed portion of the exhaust pipe and provided with a number of discharge openings. (auth)

In-situ measurement of response time of RTDs and pressure transmitters in nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.; Riner, J.L.

1993-01-01

Response time measurements are performed once every fuel cycle on most safety-related temperature and pressure sensors in a majority of nuclear power plants in the US. This paper provides a review of the methods that are used for these measurements. The methods are referred to as the Loop Current Step Response (LCSR) test, which is used for response time testing of temperature sensors, and noise analysis and power interrupt (PI) tests, which are used for response time testing of pressure, level, and flow transmitters

The online sealing performance test of the primary circuit pressure boundary check valve in nuclear power plants

International Nuclear Information System (INIS)

Yang Yunfei; Huang Huimin

2013-01-01

The primary circuit pressure boundary check valves of 320 MW pressurized water reactor is a nuclear grade I key equipment. The sealing demand is very high, which is directly related to the internal leakage rate of the primary circuit system. After the welding check valve is repaired, the sealing performance is judged by color printing checks. If there is water or humid vapor in the pipe, it will affect the accuracy of the color printing checks. For the particularity of the online check valve tightness test, online detecting device is designed by the hydraulic pressure drop method in other nuclear power plants, but the method has some shortcomings and restrictions. In this paper, we design a reliable and portable test equipment by the low-pressure gas seal test flow measurement, which make accurate and quantitative judgment of sealing property after the pressure boundary check valves are repaired. (authors)

Slovenske elektrarne, a.s., Mochovce Nuclear Power Plant

International Nuclear Information System (INIS)

1998-01-01

In this booklet the uranium atom nucleus fission as well as electricity generation in a nuclear power plant (primary circuit, reactor, reactor pressure vessel, fuel assembly, control rod and reactor power control) are explained. Scheme of electricity generation in nuclear power plant and Cross-section of Mochovce Nuclear Power Plant unit are included. In next part a reactor scram, refuelling of fuel, instrumentation and control system as well as principles of nuclear safety and safety improvements are are described

Requirements for class 1, 2, and 3 pressure-retaining systems and components in CANDU nuclear power plants

International Nuclear Information System (INIS)

1991-09-01

This third edition of CAN/CSA-N285.1 supersedes the 1981 and 1975 editions. It provides the specific requirements for design, fabrication, and installation of Class 1, 2 and 3 pressure-retaining systems and components in CANDU nuclear power plants, and over pressure protection of the heat transport system. The general requirements for pressure-retaining systems and components are given in CSA Standard CAN/CSA-N285.0, with which Class 1, 2 and 3 systems and components must also comply

Guidelines for Application of the Master Curve Approach to Reactor Pressure Vessel Integrity in Nuclear Power Plants

International Nuclear Information System (INIS)

Lyssakov, V.N.; Kang, K.S.

2005-01-01

These guidelines have been developed under an International Atomic Energy Agency (IAEA) Co-ordinated Research Project (CRP) titled ''Surveillance Programme Results Application to Reactor Pressure Vessel Integrity Assessment.'' The IAEA has sponsored a series of five CRPs that have led to a focus on measuring the best irradiation fracture parameters using relatively small test specimens for assuring structural integrity of reactor pressure vessel (RPV) materials in Nuclear Power Plants (NPPs)

Blow-off device for limiting excess pressure in nuclear power plants, especially in boiling water nuclear power plants

International Nuclear Information System (INIS)

Simon, U.; Werner, K.D.; Hoffmann, D.; Pontani, B.

1979-01-01

In a blow-off device for limiting excess pressure in nuclear power plants, at least one condensation tube disposed so that a lower outflow and thereof is immersed in a volume of water, and an upper inflow end of the condensation tube extends out of the volume of water and is connectible to a source of steam that is to be condensed or a steam-air mixture, the outflow end of the condensation tube, for stabilizing the condensation being provided with an assembly of wall parts forming passageways extending in axial direction for subdividing the steam flow and bubbles produced in the volume of water, the passageways of the assembly of wall parts being stepped in axial direction at both axial ends of the assembly of wall parts

Effect of special features of nuclear power plants

International Nuclear Information System (INIS)

Scharf, H.

1986-01-01

Special features of nuclear power plants are reported with the Muelheim-Kaerlich pressurized water reactor as the reference plant. This nuclear reactor uses 'Once Through Steam Generators (OTSG)' with 'Integrated Economizer' to provide the turbine with superheated steam. The implementation of OTSG allows to operate the plant with constant steam pressure over the entire power range, and with constant main coolant temperature over a power range from 15% power to 100% power. Control of the plant during power operation is provided by the 'Integrated Control System', which simultaneously sends signals to the plant's subsystems reactor, OTSG, and turbine to get optimum response of the plant during power transients. The characteristics of this 'Integrated Control System' and its different modes of operation are presented. (orig./GL)

1984 availability of the world's nuclear power plants

International Nuclear Information System (INIS)

Szeless, A.; Oszuszky, F.

1985-01-01

This survey of the availability of the world's nuclear power plants in 1984 coveres 250 units (the CMEA countries excluded) with an aggregate 184,500 MWe, which are arranged by types of reactor and geographic distribution. The utilization of nuclear power plant capacity attained an average of 66% in 1984, which is an increase by 3 percentage points in the utilization of capacity over the previous year's level (63%). Capacity utilization in pressurized water reactors 1984 (69%) was 4 percentage points higher than it was in boiling water reactors (65%). The ranking list of the world's nuclear power plants is headed by one heavy water reactor and one gas cooled reactor each (98%), followed by five generating units, i.e., two pressurized water reactors, two boiling water reactors, and one heavy water reactor (all 95%). The best German nuclear power plants were Grafenrheinfeld (89%, position No. 20) and Unterweser and Stade (88%, position No. 25). (orig.) [de

Preliminary development of an integrated approach to the evaluation of pressurized thermal shock as applied to the Oconee Unit 1 Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Burns, T J; Cheverton, R D; Flanagan, G F; White, J D; Ball, D G; Lamonica, L B; Olson, R

1986-05-01

An evaluation of the risk to the Oconee-1 nuclear plant due to pressurized thermal shock (PTS) has been Completed by Oak Ridge National Laboratory (ORNL). This evaluaion was part of a Nuclear Regulatory Commission (NRC) program designed to study the PTS risk to three nuclear plants: Oconee-1, a Babcock and Wilco reactor plant owned and operated by Duke Power Company; Calvert Cliffs-1, a Combustion Engineering reactor plant owned and operated by Baltimore Gas and Electric company; and H.B. Robinson-2, a Westinghouse reactor plant owned and operated by Carolina Power and Light Company. Studies of Calvert Cliffs-1 and H.B. Robinson-2 are still underway. The specific objectives of the Oconee-1 study were to: (1) provide a best estimate of the probability of a through-the-wall crack (TWC) occurring in the reactor pressure vessel as a result of PTS; (2) determine dominant accident sequences, plant features, operator and control actions and uncertainty in the PTS risk; and (3) evaluate effectiveness of potential corrective measures.

Water cooled type nuclear power plant

International Nuclear Information System (INIS)

Arai, Shigeki.

1981-01-01

Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

On-line testing of response time and calibration of temperature and pressure sensors in nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.

1995-01-01

Periodic calibrations and response time measurements are necessary for temperature and pressure sensors in the safety systems of nuclear power plants. Conventional measurement methods require the test to be performed at the sensor location or involve removing the sensor from the process and performing the tests in a laboratory or on the bench. The conventional methods are time consuming and have the potential of causing wear and tear on the equipment, can expose the test personnel to radiation and other harsh environments, and increase the length of the plant outage. Also, the conventional methods do not account for the installation effects which may have an influence on sensor performance. On-line testing methods alleviate these problems by providing remote sensor response time and calibration capabilities. For temperature sensors such as Resistance Temperature Detectors (RTDs) and thermocouples, an on-line test method called the Loop Current Step Response (LCSR) technique has been developed, and for pressure transmitters, an on-line method called noise analysis which was available for reactor diagnostics was validated for response time testing applications. Both the LCSR and noise analysis tests are performed periodically in U.S. nuclear power plants to meet the plant technical specification requirements for response time testing of safety-related sensors. Automated testing of the calibration of both temperature and pressure sensors can be accomplished through an on-line monitoring system installed in the plant. The system monitors the DC output of the sensors over the fuel cycle to determine if any calibration drift has occurred. Changes in calibration can be detected using signal averaging and intercomparison methods and analytical redundancy techniques. (author)

Thermal coupling system analysis of a nuclear desalination plant

International Nuclear Information System (INIS)

Adak, A.K.; Srivastava, V.K.; Tewari, P.K.

2010-01-01

When a nuclear reactor is used to supply steam for desalination plant, the method of coupling has a significant technical and economic impact. The exact method of coupling depends upon the type of reactor and type of desalination plant. As a part of Nuclear Desalination Demonstration Project (NDDP), BARC has successfully commissioned a 4500 m 3 /day MSF desalination plant coupled to Madras Atomic Power Station (MAPS) at Kalpakkam. Desalination plant coupled to nuclear power plant of Pressurized Heavy Water Reactor (PHWR) type is a good example of dual-purpose nuclear desalination plant. This paper presents the thermal coupling system analysis of this plant along with technical and safety aspects. (author)

Studies on the effectiveness of measures to maintain the integrity of pressurized components in German nuclear power plants. Final report

International Nuclear Information System (INIS)

Elmas, M.; Jendrich, U.; Michel, F.; Reck, H.; Schimpfke, T.; Walter, M.; Wenke, R.

2013-03-01

The overall objective of the project was to investigate the effectiveness of measures to maintain the as-built quality of the pressure-retaining components in German nuclear power plants. In particular, investigations were performed on the application of the break preclusion concept, existing monitoring systems and the significance of the pressure test as part of the inspection concept. Moreover, the KompInt knowledge base has been updated. Break preclusion for pipes was applied in all German plants already during planning or after commissioning to a varying extent. The basic features of the required assessments were considered in the German nuclear regulations for the first time by inclusion in the safety requirements for nuclear power plants of 2012. The requirements for assessments, differing in their degree of detail, in the interpretations of these safety requirements and in the safety standard KTA 3206 are still in the draft stage. For the first time, the vessels as well as housings of valves and pumps are also included in the concept. Through the use of advanced monitoring systems it was possible in German plants at an early stage to establish modes of operation that minimise the load on components, to carry out appropriate technical backfitting measures, and to identify damages. In plant areas where local water chemistry parameters may result that deviate from the specification, the effectiveness of water chemistry monitoring is limited. In this case, other operational measures must be taken. The results of the simulations performed with the help of the GRS-developed PROST computer code to determine the significance of pressure tests lead - in accordance with the results of operating experience evaluation - to the conclusion that pressure tests carried out within the pressure-retaining boundary contribute to safeguarding the integrity. The user-friendliness of the KompInt knowledge base has been increased by changing over to a new hardware, a software

Assessment of missiles generated by pressure component failure and its application to recent gas-cooled nuclear plant design

International Nuclear Information System (INIS)

Tulacz, J.; Smith, R.E.

1980-01-01

Methods for establishing characteristics of missiles following pressure barrier rupture have been reviewed in order to enable evaluation of structural response to missile impact and to aid the design of barriers to protect essential plant on gas cooled nuclear plant against unacceptable damage from missile impact. Methods for determining structural response of concrete barriers to missile impact have been reviewed and some methods used for assessing the adequacy of steel barriers on gas-cooled nuclear plant have been described. The possibility of making an incredibility case for some of the worst missiles based on probability arguments is briefly discussed. It is shown that there may be scope for such arguments but there are difficulties in quantifying some of the probability factors. (U.K.)

Summary of nuclear power plant construction

International Nuclear Information System (INIS)

Tamura, Saburo

1973-01-01

Various conditions for the construction of nuclear power plants in Japan without natural resources were investigated. Expansion of the sites of plants, change of reactor vessels, standardization of nuclear power plants, possiblity of the reduction of construction period, approaching of nuclear power plants to consuming cities, and group construction were studied. Evaluation points were safety and economy. Previous sites of nuclear power plants were mostly on plane ground or cut and enlarge sites. Proposals for underground or offshore plants have been made. The underground plants were made at several places in Europe, and the ocean plant is now approved in U.S.A. as a plant on a man-made island. Vessels for containing nuclear reactors are the last barriers to the leakage of radioactive substance. At the initial period, the vessels were made of steel, which were surrounded by shielding material. Those were dry well type containers. Then, vessel type changed to pressure-suppression type wet containers. Now, it tends to concrete (PC or RC) type containers. There is the policy on the standardization of nuclear power plants by U.S.A.E.C. in recent remarkable activity. The merit and effect of the standardization were studied, and are presented in this paper. Cost of the construction of nuclear power plants is expensive, and interest of money is large. Then, the reduction of construction period is an important problem. The situations of plants approaching to consuming cities in various countries were studied. Idea of group construction is described. (Kato, T.)

Nuclear power plant operating experience, 1976

International Nuclear Information System (INIS)

1977-11-01

This report is the third in a series of reports issued annually that summarize the operating experience of U.S. nuclear power plants in commercial operation. Power generation statistics, plant outages, reportable occurrences, fuel element performance, occupational radiation exposure and radioactive effluents for each plant are presented. Summary highlights of these areas are discussed. The report includes 1976 data from 55 plants--23 boiling water reactor plants and 32 pressurized water reactor plants

Estimation of residual stress distribution for pressurizer nozzle of Kori nuclear power plant considering safe end

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-08-15

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which considered safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

Requirements for class 1C, 2C, and 3C pressure-retaining components and supports in CANDU nuclear power plants

International Nuclear Information System (INIS)

1989-01-01

This Standard applies to pressure-retaining components of CANDU nuclear power plants that have a code classification of Class 1C, 2C or 3C. These are pressure-retaining components where, because of the design concept, the rules of the ASME Boiler and Pressure Vessel Code do not exist, are not applicable, or are not sufficient. The Standard provides rules for the design, fabrication, installation, examination and inspection of these components and supports. It provides rules intended to ensure the pressure-retaining integrity of components, not the operability. It also provides rules for the support of fueling machines. The Standard applies only to new construction prior to the plant being declared in service

Poison and diluent system for nuclear power plants

International Nuclear Information System (INIS)

Parker, W.G.; Ravets, J.M.; Preble, B.S.

1978-01-01

A system to prevent supercriticality in nuclear power plants in the unlikely event of a core destructive accident terminating in the nuclear core meltdown is described. The system dilutes and poisons the molten core to maintain subcriticality, and is useful in mobile nuclear power plants, or in nuclear plants subject to seismic disturbances, where the orientation of the nuclear reactor after the accident is unknown. It is also applicable to alleviate the consequences of loss of coolant flow accidents from any cause. Aside from preventing supercriticality, the system serves the dual purpose of acting as a biological shield and/or structural member that reduces the deleterious effects of accidental core impaction, without compromising power plant weight and size constraints. A borated material, with a melting point greater than the fuel melting point, is inserted in the pressure vessel behind an inner wall. In the unlikely event of a core meltdown, the molten fuel melts through the inner wall and is diluted and poisoned by the borated material. In the event the molten fuel melts through the pressure vessel, additional borated material is provided to continue diluting and poisoning

Inspection of Nuclear Power Plant Structures - Overview of Methods and Related Applications

International Nuclear Information System (INIS)

Naus, Dan J.

2009-01-01

The objectives of this limited study were to provide an overview of the methods that are available for inspection of nuclear power plant reinforced concrete and metallic structures, and to provide an assessment of the status of methods that address inspection of thick, heavily-reinforced concrete and inaccessible areas of the containment metallic pressure boundary. In meeting these objectives a general description of nuclear power plant safety-related structures was provided as well as identification of potential degradation factors, testing and inspection requirements, and operating experience; methods for inspection of nuclear power plant reinforced concrete structures and containment metallic pressure boundaries were identified and described; and applications of nondestructive evaluation methods specifically related to inspection of thick-section reinforced concrete structures and inaccessible portions of containment metallic pressure boundaries were summarized. Recommendations are provided on utilization of test article(s) to further advance nondestructive evaluation methods related to thick-section, heavily-reinforced concrete and inaccessible portions of the metallic pressure boundary representative of nuclear power plant containments. Conduct of a workshop to provide an update on applications and needed developments for nondestructive evaluation of nuclear power plant structures would also be of benefit.

Inspection of Nuclear Power Plant Structures - Overview of Methods and Related Applications

Energy Technology Data Exchange (ETDEWEB)

Naus, Dan J [ORNL

2009-05-01

The objectives of this limited study were to provide an overview of the methods that are available for inspection of nuclear power plant reinforced concrete and metallic structures, and to provide an assessment of the status of methods that address inspection of thick, heavily-reinforced concrete and inaccessible areas of the containment metallic pressure boundary. In meeting these objectives a general description of nuclear power plant safety-related structures was provided as well as identification of potential degradation factors, testing and inspection requirements, and operating experience; methods for inspection of nuclear power plant reinforced concrete structures and containment metallic pressure boundaries were identified and described; and applications of nondestructive evaluation methods specifically related to inspection of thick-section reinforced concrete structures and inaccessible portions of containment metallic pressure boundaries were summarized. Recommendations are provided on utilization of test article(s) to further advance nondestructive evaluation methods related to thick-section, heavily-reinforced concrete and inaccessible portions of the metallic pressure boundary representative of nuclear power plant containments. Conduct of a workshop to provide an update on applications and needed developments for nondestructive evaluation of nuclear power plant structures would also be of benefit.

Maintenance of process instrumentation in nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.

2006-01-01

The resurgence of the nuclear power industry at a time when the nuclear plant I and C workforce is diminishing due to aging and retirements has been the motivation for writing this book. This book compiles 30 years of practical knowledge gained by the author and his staff in testing the I and C systems of nuclear power plants around the world. It focuses on process temperature and pressure sensors and the verification of these sensors' calibration and response time. In spite of great advances in electronics, computers, and measurement technologies, important process parameters such as temperature and pressure are still measured with conventional sensing techniques found in RTDs, thermocouples, and conventional pressure and differential pressure sensors. Furthermore, no improved technology providing comparable performance is currently on the horizon. Therefore, it is important to understand how these sensors function and the testing techniques for verifying their performance. This book is intended to help provide this understanding. (orig.)

Ventilation-air conditioner system in nuclear power plant

International Nuclear Information System (INIS)

Kubota, Ryuji; Sugisaki, Toshihiko.

1989-01-01

This invention concerns a ventilation-air conditioner system which enables, upon occurrence of accidents in a nuclear power plant, continuous operation for other adjacent nuclear power plants with no effect of accidents. Air supply system and exhaust system are operated during usual operaiton. If loss of coolants accidents should occur in an adjacent nuclear power plants, operation is switched from ventilation operaiton to the operation of re-cycling system based on an AND logic of three signals, that is, a pressure HIGH signal for the reactor container, a water level LOW signal for the reactor and a radioactivity signal of the ventilation-air conditioner sytem on the side of air supply in the nuclear power plant. Thus, nuclear reactor buildings of the nuclear power plant are from the external atmosphere. Therefore, the radioactivity HIGH signal for switching to the emergency air conditioner system of the nuclear power plant is not actuated due to the loss of coolant accidents in the adjacent nuclear power plant. In addition, since the atmospheric temperature in the nuclear reactor building can be maintained by a cooling device disposed to the recycling system, reactor shutdown can be prevented. (I.S.)

HRA qualitative analysis in a nuclear power plant

International Nuclear Information System (INIS)

Dai Licao; Zhang Li; Huang Shudong

2004-01-01

Human reliability analysis (HRA) is a very important part of probability safety assessment (PSA) in a nuclear power plant. Qualitative analysis is the basis and starting point of HRA. The purpose, the principle, the method and the procedure of qualitative HRA are introduced. SGTR, a pressurized nuclear power plant as an example, is used to illustrate it. (authors)

77 FR 47121 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Units 1 and 2...

Science.gov (United States)

2012-08-07

... for Nuclear Power Plant Personnel,'' endorses the Nuclear Energy Institute (NEI) report NEI 06-11...(c)(25). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment... hereafter in effect. The facility consists of two pressurized-water reactors (PWRs) located in Calvert...

Pressure fluctuation analysis for charging pump of chemical and volume control system of nuclear power plant

Directory of Open Access Journals (Sweden)

Chen Qiang

2016-01-01

Full Text Available Equipment Failure Root Cause Analysis (ERCA methodology is employed in this paper to investigate the root cause for charging pump’s pressure fluctuation of chemical and volume control system (RCV in pressurized water reactor (PWR nuclear power plant. RCA project task group has been set up at the beginning of the analysis process. The possible failure modes are listed according to the characteristics of charging pump’s actual pressure fluctuation and maintenance experience during the analysis process. And the failure modes are analysed in proper sequence by the evidence-collecting. It suggests that the gradually untightened and loosed shaft nut in service should be the root cause. And corresponding corrective actions are put forward in details.

Pumps in nuclear power plants

International Nuclear Information System (INIS)

Kim, J.H.

1991-01-01

This paper reports that pumps play an important role in nuclear plant operation. For instance, reactor coolant pumps (RCPs) should provide adequate cooling for reactor core in both normal operation and transient or accident conditions. Pumps such as Low Pressure Safety Injection (LPSI) pump in the Emergency Core Cooling System (ECCS) play a crucial role during an accident, and their reliability is of paramount importance. Some key issues involved with pumps in nuclear plant system include the performance of RCP under two-phase flow conditions, piping vibration due to pump operating in two-phase flows, and reliability of LPSI pumps

Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

International Nuclear Information System (INIS)

Sykora, D.

1986-01-01

An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

Friction and wear studies of nuclear power plant components in pressurized high temperature water environments

International Nuclear Information System (INIS)

Ko, P.L.; Zbinden, M.; Taponat, M.C.; Robertson, M.F.

1997-01-01

The present paper is part of a series of papers aiming to present the friction and wear results of a collaborative study on nuclear power plant components tested in pressurized high temperature water. The high temperature test facilities and the methodology in presenting the kinetics and wear results are described in detail. The results of the same material combinations obtained from two very different high temperature test facilities (NRCC and EDF) are presented and discussed. (K.A.)

Safety in nuclear power plants

International Nuclear Information System (INIS)

Koeberlein, K.

1987-01-01

In nuclear power plants large amounts of radioactive fission products ensue from the fission of uranium. In order to protect the environment, the radioactive material is confined in multiple 'activity barriers' (crystal matrix of the fuel, fuel cladding, coolant boundary, safety containment, reactor building). These barriers are protected by applying a defense-in-depth concept (high quality requirements, protection systems which recognize and terminate operational incidents, safety systems to cope with accidents). In spite of a favorable safety record of German nuclear power plants it is obvious - and became most evident by the Chernobyl accident - that absolute safety is not achievable. At Chernobyl, however, design disadvantages of that reactor type (like positive reactivity feedback of coolant voiding, missing safety containment) played an important role in accident initiation and progression. Such features of the Russian 'graphite-moderated pressure tube boiling water reactor' are different from those of light water reactors operating in western countries. The essential steps of the waste management of the nuclear fuel cycle ('Entsorgung') are the interim storage, the shipment, and the reprocessing of the spent fuel and the final repository of radioactive waste. Reprocessing means the separation of fossil material (uranium, plutonium) from radioactive waste. Legal requirements for radiological protection of the environment, which are identical for nuclear power plants and reprocessing plant, are complied with by means of comprehensive filter systems. Safety problems of a reprocessing plant are eased considerably by the fact that system pressures, process temperatures and energy densities are low. In order to confine the radioactive waste from the biosphere for a very long period of time, it is to be discarded after appropriate treatment into the deep geological underground of salt domes. (orig./HP) [de

Nuclear power plants

International Nuclear Information System (INIS)

1985-01-01

Data concerning the existing nuclear power plants in the world are presented. The data was retrieved from the SIEN (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: nuclear plants, its status and type; installed nuclear power plants by country; nuclear power plants under construction by country; planned nuclear power plants by country; cancelled nuclear power plants by country; shut-down nuclear power plants by country. (E.G.) [pt

Method of operating a nuclear turbine plant

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Hiraku; Ootawara, Yasuhiko; Imai, Tetsu

1985-04-25

A method is presented to prevent the lowering in the reactor feedwater temperature thereby secure necessary amount of steams even in a plant operation under low load. The feedwater temperature of a nuclear reactor is detected at the low load region of the plant and high enthalpy steams are supplied to a high pressure feedwater heater by opening a supply stream extract switching valve. This enables to maintain the feedwater temperature in the nuclear reactor at a constant level.

Assessment of fiber optic sensors and other advanced sensing technologies for nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.

1996-01-01

As a result of problems such as calibration drift in nuclear plant pressure sensors and the recent oil loss syndrome in some models of Rosemount pressure transmitters, the nuclear industry has become interested in fiber optic pressure sensors. Fiber optic sensing technologies have been considered for the development of advanced instrumentation and control (I ampersand C) systems for the next generation of reactors and in older plants which are retrofitted with new I ampersand C systems. This paper presents the results of a six-month Phase I study to establish the state-of-the-art in fiber optic pressure sensing. This study involved a literature review, contact with experts in the field, an industrial survey, a site visit to a fiber optic sensor manufacturer, and laboratory testing of a fiber optic pressure sensor. The laboratory work involved both static and dynamic performance tests. This initial Phase I study has recently been granted a two-year extension by the U.S. Nuclear Regulatory Commission (NRC). The next phase will evaluate fiber optic pressure sensors in specific nuclear plant applications in addition to other advanced methods for monitoring critical nuclear plant equipment

Introduction to Exxon nuclear fuel fabrication plant

International Nuclear Information System (INIS)

Schneider, R.A.

1985-01-01

The Exxon Nuclear low-enriched uranium fuel fabrication plant in Richland, Washington produces fuel assemblies for both pressurized water and boiling water reactors. The Richland plant was the first US bulk-handling facility selected by the IAEA for inspection under the US-IAEA Safeguards Agreement. The plant was under IAEA inspection from March 1981 through October 1983. This text provides a written description of the plant layout, operation and process. The text also includes a one ton-a-day model (or reference) plant which was adapted from the Exxon Nuclear plant. The Model Plant provides a generic example of a low-enriched uranium (LEU) bulk-handling facility. The Model Plant is used to illustrate in a more quantitative way some of the key safeguards requirements for a bulk-handling facility

Application of environmentally-corrected fatigue curves to nuclear power plant components

International Nuclear Information System (INIS)

Ware, A.G.; Morton, D.K.; Nitzel, M.E.

1996-01-01

Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four US nuclear steam supply system vendors. For each facility, six locations were studied including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This paper discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

Safety in Swiss nuclear power plants

International Nuclear Information System (INIS)

Cederqvist, H.

1992-01-01

Safety-related facilities and equipment are continuously backfitted in Swiss nuclear power plants. In the Beznau-1 and -2 nuclear generating units, the measures taken under the heading of 'Backfitting of Emergency Systems' included provisions to enhance the protection against earthquakes, airplane crash, and fire; in addition, the emergency power system was upgraded. In Muehleberg, the stack exhaust air monitoring system was optimized. The containment pressure suppression system of the plant has been designed to withstand a hypothetical accident exceeding the design basis. The BKM-Crud computer simulation model simulates steps taken to reduce radiation exposure. The power of Swiss nuclear power stations will be raised by 4% to 15% within the 'Energy 2000' action program. (orig.) [de

CAS medium-size nuclear plants

International Nuclear Information System (INIS)

Vogelweith, L.; Weiss, A.

1977-01-01

CEA has developed a range of pressurized water reactors of the type CAS Compact, which are adapted to civil ship propulsion, or to electric power production, combined possibly with heat production, up to outputs equivalent to 125MW(e). Nuclear plants equipped with these reactors are suitable for medium-size electric networks, especially in developing countries, because they are easily adaptable, owing to their flexibility; they can be installed and used in a variety of ways (on land, floating installation, combination of electric power and other production, etc.); they can be used as training reactors by countries wishing to limit their investment plans before undertaking a wider nuclear development. Examples of two possible realizations are presented: as a floating plant, and as a combined electric and desalting plant. (author)

WWER-440/230 reactor pressure vessel integrity. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-08-01

This report was prepared with the objective of integrating all aspects involved and to provide plant specific information on the issue of reactor pressure vessel integrity including pressurized thermal shock assessment. Areas of the thermal hydraulic analysis including selection of transients, of the structural analysis including fracture mechanics assessment and of the material properties including embrittlement, annealing and re-embrittlement behaviour are addressed. The report also provides related recommendations and conclusions as well as detailed information on the plant specific status for operating WWER-440/230 nuclear power plants. 10 refs, 9 figs, 9 tabs

Nuclear power plant containment metallic pressure boundary materials and plans for collecting and presenting their properties

International Nuclear Information System (INIS)

Oland, C.B.

1995-04-01

A program is being conducted at the Oak Ridge National Laboratory (ORNL to assist the Nuclear Regulatory Commission (NRC)) in their assessment of the effects of degradation (primarily corrosion) on the structural capacity and leaktight integrity of metal containments and steel liners of reinforced concrete structures in nuclear power plants. One of the program objectives is to characterize and quantify manifestations of corrosion on the properties of steels used to construct containment pressure boundary components. This report describes a plan for use in collecting and presenting data and information on ferrous alloys permitted for use in construction of pressure retaining components in concrete and metal containments. Discussions about various degradation mechanisms that could potentially affect the mechanical properties of these materials are also included. Conclusions and recommendations presented in this report will be used to guide the collection of data and information that will be used to prepare a material properties data base for containment steels

Activity determination for neutron dosimetry in the vigilance programme for the pressure vessel in Atucha I nuclear power plant

International Nuclear Information System (INIS)

Furnari, J.C.; Cohen, I.M.; Ciriani, D.F.; Helzel Garcia, J.

1993-01-01

The methodologies for the activity determination of Co-60, Nb-93m and Nb-94 in flux monitors are presented. This was done in order to evaluate dose and damage caused by radiation received by pressure vessel materials of the Atucha I nuclear power plant for its surveillance program. (author)

Coupling technology for dual-purpose nuclear-desalting plants

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Anderson, T.D.; Reed, S.A.

1976-11-01

Although the basic technology for the various components of nuclear dual-purpose plants is reasonably well developed, the techniques of coupling the elements together to form a reliable, economical system that will satisfy the diverse operating requirements are not well established. The purpose of the study reported is to examine the technical, economic, and safety considerations in coupling nuclear power plants with distillation units to form a dual-purpose power and water distillation plant. The basic coupling arrangement required to provide a large-scale dual-purpose water plant is to supply steam to the water plant from the exhaust of a back-pressure turbine. The principal component at the interface that may require major research and development is the back-pressure turbine. To satisfy the operational requirements, two major auxiliary systems will be needed. These are: (1) a prime-steam bypass system, and (2) auxiliary condensers. These systems will provide a degree of independence between water and power production and can be justified economically

Major plant retrofits at Monticello nuclear generating plant

International Nuclear Information System (INIS)

Larsen, D.E.; Hogg, C.B.

1986-01-01

For the past several years, Northern States Power (NSP) has been making major plant retrofits to Monticello Nuclear generating Station in order to improve plant availability and upgrade the plant components for the potential extension of the operating license (life extension). This paper discusses in detail three major retrofits that have been completed or in the process of completion; recirculation loop piping replacement, reactor pressure vessel (RPV) water level-instrumentation modification, core spray piping replacement, the authors will address the scope of work, design and installation concerns, and life extension considerations during the design and procurement process for these three projects

SCW Pressure-Channel Nuclear Reactor Some Design Features

Science.gov (United States)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

Science.gov (United States)

2013-01-22

... Analysis Reports for Nuclear Power Plants: LWR Edition.'' The new subsection is the Standard Review Plan... Nuclear Power Plants: Integral Pressurized Water Reactor (iPWR) Edition.'' DATES: Comments must be filed... NUCLEAR REGULATORY COMMISSION [NRC-2012-0268] Review of Safety Analysis Reports for Nuclear Power...

New method of safety assessment for pressure vessel of nuclear power plant--brief introduction of master curve approach

International Nuclear Information System (INIS)

Yang Wendou

2011-01-01

The new Master Curve Method is called as a revolutionary advance to the assessment of- reactor pressure vessel integrity in USA. This paper explains the origin, basis and standard of the Master Curve from the reactor pressure-temperature limit curve which assures the safety of nuclear power plant. According to the characteristics of brittle fracture which is greatly susceptible to the microstructure, the theory and the test method of the Master Curve as well as its statistical law which can be modeled using Weibull distribution are described in this paper. The meaning, advantage, application and importance of the Master Curve as well as the relation between the Master Curve and nuclear power safety are understood from the fitting formula for the fracture toughness database by Weibull distribution model. (author)

Medium-size nuclear plants

International Nuclear Information System (INIS)

Vogelweith, L.; Lavergne, J.C.; Martinot, G.; Weiss, A.

1977-01-01

CEA (TECHNICATOME) has developed a range of pressurized water reactors of the type ''CAS compact'' which are adapted to civil ship propulsion, or to electric power production, combined possibly with heat production, up to outputs equivalent to 125 MWe. Nuclear plants equipped with these reactors are suitable to medium-size electric networks. Among the possible realizations, two types of plants are mentioned as examples: 1) Floating electron-nuclear plants; and 2) Combined electric power and desalting plants. The report describes the design characteristics of the different parts of a 125 MWe unit floating electro-nuclear plant: nuclear steam system CAS 3 G, power generating plant, floating platform for the whole plant. The report gives attention to the different possibilities according to site conditions (the plant can be kept floating, in a natural or artificial basin, it can be put aground, ...) and to safety and environment factors. Such unit can be used in places where there is a growing demand in electric power and fresh water. The report describes how the reactor, the power generating plant and multiflash distillation units of an electric power-desalting plant can be combined: choice of the ratio water output/electric power output, thermal cycle combination, choice of the gain ratio, according to economic considerations, and to desired goal of water output. The report analyses also some technical options, such as: choice of the extraction point of steam used as heat supply of the desalting station (bleeding a condensation turbine, or recovering steam at the exhaust of a backpressure turbine), design making the system safe. Lastly, economic considerations are dealt with: combining the production of fresh water and electric power provides usually a much better energy balance and a lower cost for both products. Examples are given of some types of installations which combine medium-size reactors with fresh water stations yielding from 10000 to 120000 m 3 per day

Improving operator quality at Genkai Nuclear Power Plant

International Nuclear Information System (INIS)

Kuwano, Takeshi.

1989-01-01

Improvement in operator quality, or improvement in an operator's skill and professional knowledge, is of prime importance because of its great influence on safe and steady plant operation. This paper describes the education and training of reactor operators at the Genkai pressurized water reactor nuclear power plant with respect to the following topics: organization of Genkai power plant; education and training program management; training at the Nuclear Training Center; training and education on-site including emergency procedures training, normal operating procedures training, informational study of emergency conditions in existing plants, and all-around training of operators; qualifying tests for supervisors; and operator motivation

AP1000 plant construction in China: Ansaldo Nucleare contribution

International Nuclear Information System (INIS)

Frogheri, Monica; Saiu, Gianfranco

2009-01-01

On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract includes: the design of the steel containment vessel, preparation of construction and fabrication, specifications, design and supply of SCV mechanical penetrations, air locks and equipment hatches. Moreover, Ansaldo Nucleare is in charge of the final design of the AP1000 PRHR-HX and together with Mangiarotti Nuclear will supply the component for the Sanmen Unit 1 NPP. The paper presents an overview of the design and manufacturing activities performed by Ansaldo Nucleare and its partners for the AP1000 plant in China. (authors)

Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

International Nuclear Information System (INIS)

Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon

2012-01-01

Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

Simulating the steam generator and the pressurizer of a PWR nuclear power plant

International Nuclear Information System (INIS)

De Greef, J.F.

1985-01-01

In a PWR nuclear power plant, considered as a power generating device, the steam generator as a subset plays an important role in the generation process, whereas the pressurizer rather acts as a control device for security purposes. Nevertheless, from a thermodynamical point of view, the two subsets behave basically in the same way, so that a common set of basic equations may be suggested to develop for each the proper mathematical simulation model. In this paper the generation of this common set of basic equations is described, from which a specific model for each device is derived. A numerical illustration of the behaviour of the two devices for typical inputs to the derived simulation model is pictured. (author)

SECURE nuclear district heating plant

International Nuclear Information System (INIS)

Nilsson; Hannus, M.

1978-01-01

The role foreseen for the SECURE (Safe Environmentally Clean Urban REactor) nuclear district heating plant is to provide the baseload heating needs of primarily the larger and medium size urban centers that are outside the range of waste heat supply from conventional nuclear power stations. The rationale of the SECURE concept is that the simplicity in design and the inherent safety advantages due to the use of low temperatures and pressures should make such reactors economically feasible in much smaller unit sizes than nuclear power reactors and should make their urban location possible. It is felt that the present design should be safe enough to make urban underground location possible without restriction according to any criteria based on actual risk evaluation. From the environmental point of view, this is a municipal heat supply plant with negligible pollution. Waste heat is negligible, gaseous radioactivity release is negligible, and there is no liquid radwaste release. Economic comparisons show that the SECURE plant is competitive with current fossil-fueled alternatives. Expected future increase in energy raw material prices will lead to additional energy cost advantages to the SECURE plant

Extended fuel cycle operation for pressurized water reactor plants

International Nuclear Information System (INIS)

Silvestri, G.J. Jr.

1978-01-01

A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

Innovative configuration of a hybrid nuclear-solar tower power plant

International Nuclear Information System (INIS)

Popov, Dimityr; Borissova, Ana

2017-01-01

This paper proposes a combination of a nuclear and a CSP plant and performs a thermodynamic analysis of the potential benefit. Most of today's operating nuclear reactor systems are producing saturated steam at relatively low pressure. This, in turn, limits their thermodynamic efficiency. Superheating of nuclear steam with solar thermal energy has the potential to overcome this drawback. Accordingly, an innovative configuration of a hybrid nuclear-CSP plant is assembled and simulated. It brings together pressurized water reactor and solar tower. The solar heat is transferred to nuclear steam to raise its temperature. Continuous superheating is provided through thermal energy storage. The results from design point calculations show that solar superheating has the potential to increase nuclear plant electric efficiency significantly, pushing it to around 37.5%. Solar heat to electricity conversion efficiency reaches unprecedented rates of 56.2%, approaching the effectiveness of the modern combined cycle gas turbine plants. Off-design model was used to simulate 24-h operation for one year by simulating 8760 cases. Due to implementation of thermal energy storage non-stop operation is manageable. The increased efficiency leads to solar tower island installed cost reductions of up to 25% compared to the standalone CSP plant, particularly driven by the smaller solar field. - Highlights: • External superheating of nuclear steam with solar thermal energy is proposed. • Novel hybrid plant configuration is assembled, modeled and simulated. • Substantial increase of nuclear plant capacity and efficiency is reported. • Superior efficiency of solar heat to electricity conversion is achieved. • Substantial decrease of solar field investment cost is reported.

Master curve approach to monitor fracture toughness of reactor pressure vessels in nuclear power plants

International Nuclear Information System (INIS)

2009-10-01

A series of coordinated research projects (CRPs) have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on reactor pressure vessel (RPV) steels. The purpose of the CRPs was to develop correlative comparisons to test the uniformity of results through coordinated international research studies and data sharing. The overall scope of the eighth CRP (CRP-8), Master Curve Approach to Monitor Fracture Toughness of Reactor Pressure Vessels in Nuclear Power Plants, has evolved from previous CRPs which have focused on fracture toughness related issues. The ultimate use of embrittlement understanding is application to assure structural integrity of the RPV under current and future operation and accident conditions. The Master Curve approach for assessing the fracture toughness of a sampled irradiated material has been gaining acceptance throughout the world. This direct measurement of fracture toughness approach is technically superior to the correlative and indirect methods used in the past to assess irradiated RPV integrity. Several elements have been identified as focal points for Master Curve use: (i) limits of applicability for the Master Curve at the upper range of the transition region for loading quasi-static to dynamic/impact loading rates; (ii) effects of non-homogeneous material or changes due to environment conditions on the Master Curve, and how heterogeneity can be integrated into a more inclusive Master Curve methodology; (iii) importance of fracture mode differences and changes affect the Master Curve shape. The collected data in this report represent mostly results from non-irradiated testing, although some results from test reactor irradiations and plant surveillance programmes have been included as available. The results presented here should allow utility engineers and scientists to directly measure fracture toughness using small surveillance size specimens and apply the results using the Master Curve approach

Development of a decommissioning plan for nuclear power plant 'Krsko'

International Nuclear Information System (INIS)

Tankosic, Djurica; Fink, Kresimir

1991-01-01

Nuclear Power Plant 'Krsko' (NEK), is the only nuclear power plant in Yugoslavia, is a two-loop, Westinghouse-design, pressurized water reactor rated at 632 MWe. When NEK applied for an operating license in 1981, it did not have to explain how the plant would be decommissioned and decommissioning provisions were not part of the licensing process. Faced with mounting opposition to nuclear power and a real threat that the plant would be shut down, the plant management developed a Mission Plan for resolving the decommissioning problem. The Mission Plan calls for a preliminary decommissioning plan to be prepared and submitted to the local regulatory body before the end of 1992

Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

International Nuclear Information System (INIS)

Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G.; Souto, Joao P.R.S.; Carvalho Junior, Ideir T.

2013-01-01

Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

International Nuclear Information System (INIS)

Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

1993-01-01

The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

Changing nuclear plant operating limits during startup and shutdown

International Nuclear Information System (INIS)

Arnold, E.C.; Carlson, R.W.; Ray, N.K.; Roarty, D.H.

1990-01-01

During startup and shutdown operation of pressurized water reactor (PWR) nuclear power plants, a low pressure decay heat removal system is used to maintain core cooling. During these phases of operation, there are numerous operating practices and design limits to meet special and sometimes conflicting requirements unique to these operations. This paper evaluates the impact and interdependencies of recent issues on plant operation and design

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Development of life evaluation technology for nuclear power plant components

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Kim, Yun Jae; Choi, Jae Boong [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

2002-03-15

This project focuses on developing reliable life evaluation technology for nuclear power plant components, and is divided into two parts, development of a life evaluation system for nuclear pressure vessels and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered in this project: defect assessment method for steam generator tubes, development of fatigue monitoring system, assessment of corroded pipes, domestic round robin analysis for constructing P-T limit curve for RPV, development of probabilistic integrity assessment technique, effect of aging on strength of dissimilar welds, applicability of LBB to cast stainless steel, and development of probabilistic piping fracture mechanics.

I and C upgrading at nuclear power plants

International Nuclear Information System (INIS)

Tamiri, A.

2003-01-01

Continuing the operation of existing nuclear power plants will help reduce the number of new base-load nuclear and fossil power plants that need to be built. Old nuclear power plants in Canada are operating with analog instrumentation and control systems. For a number of reasons, such as changes and improvements in the applicable standards and design, maintenance problems due to the lack of spares, technical obsolescence, the need to increase power production, availability, reliability and safety, and in order to reduce operation and maintenance costs, instrumentation and control upgrading at nuclear power plants in a cost effective manner should be considered the greatest priority. Failures of instrumentation and control (I and C) due to aging and obsolescence issues may have an immediate negative impact on plant reliability and availability and also affect long-term plant performance and safety. In today's competitive marketplace, power plants are under pressure to cut spending on maintenance while reducing the risk of equipment failure that could cause unplanned outage. To improve plant safety and availability, old nuclear power plants will require investment in new technologies that can improve the performance and reduce the costs of generation by addressing the long term reliability of systems by up-grading to modem digital instrumentation and control and optimization opportunities. Boiler drum level control at nuclear power plants is critical for both plant protection and equipment safety and applies equality to high and low levels of water within the boiler drum. Plant outage studies at Pickering Nuclear have identified boiler drum level control and feed water control systems as major contributors to plant unavailability. Ways to improve transient and steady state response, upgrading existing poor analog control systems for boiler level and feed-water control systems at Pickering Nuclear, with enhanced and robust controller will be discussed in this paper

Pressure test at the reactor building of the Embalse Nuclear Power Plant (CNE)

International Nuclear Information System (INIS)

Coutsiers, E.E.; Perrino, J.; Moreno, C.; Batistic, J.A.; Lolis, R.R.; Aviles, A.

1991-01-01

Upon request by the Licensing Authority, the reactor building (RB) in a nuclear power plant must be submitted to pressure tests. One of these tests is to be performed before startup and, then, a test must be carried out every 5 years in operation. The pre-operational tests took place in August 1981, under two values of relative pressure: 1.266 kg/cm 2 and 0.422 kg/cm 2 . Operational tests must only be made at the lower pressure and their objective is to verify that the loss speed remains within the range indicated in the corresponding technical specification. The first operational test was performed in August 1989. The personnel of the CNE took care of the preparation of the Work Plan, of aligning the various systems contained in the RB, of pressurization, of monitoring localized tightedness, of depressurization and of the general and quality control of the test. The measurements were carried out by the CISME (Center of Metrology Research and Service) of the National Institute of Industrial Technology (INTI) , which did also supply the necesary instruments and the data collection system. There is also a description of the work performed before the test, of the calculation method used for assessing the loss rate, of the test sequencies and of the results obtained. (Author) [es

Radioactive waste processing method for a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Kuriyama, O

1976-06-04

Object is to subject radioactive liquid waste in a nuclear power plant to reverse permeation process after which it is vaporized and concentrated thereby decreasing the quantity of foam to be used to achieve effective concentration of the liquid waste. Liquid waste containing a radioactive material produced from a nuclear power plant is first applied with pressure in excess of osmotic pressure by a reverse permeation device and is separated into clean water and concentrated liquid by semi-permeable membrane. Next, the thus reverse-permeated and concentrated waste is fed to an evaporator which control foaming by the foam and then further reconcentrated for purification of the liquid waste.

Development of life evaluation technology for nuclear power plant components

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

2001-03-15

This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

Nuclear plant analyzer development and analysis applications

International Nuclear Information System (INIS)

Laats, E.T.

1984-10-01

The Nuclear Plant Analyzer (NPA) is being developed as the US Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

Directory of Open Access Journals (Sweden)

Xiangjie Liu

2014-01-01

Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

LWR nuclear power plant component failures

International Nuclear Information System (INIS)

Schmidt, W.H.

1980-10-01

An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

After-heat removing device in nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Mizuno, K [Nippon Atomic Industry Group Co. Ltd., Tokyo

1977-01-14

Purpose: To prevent water hammer in a BWR type reactor or the like by moving water in pipe lines having stagnant portions in an after-heat removing device. Constitution: To a reactor container, is provided a recycling pump which constitutes a closed loop type recycling system in a nuclear power plant together with a pressure vessel and pipe lines. A pump and a heat exchanger are provided outside of the reactor container and they are connected to up- and down-streams of the recycling pump to form an after-heat removing device in the plant. Upon shutdown of the nuclear power plant, since water in the stagnant portion flows to the intake port of the recycling pump and water from the reactor is spontaneously supplemented thereafter to the stagnant portion, neither pressurized water nor heated steam is generated and thus water hammer is prevented.

Nuclear reactor plant

International Nuclear Information System (INIS)

Schabert, H.P.; Laurer, E.

1977-01-01

The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

Training and qualification of nuclear power plant operators

International Nuclear Information System (INIS)

Ohsuga, Y.

2008-01-01

Based on training experiences of the nuclear power plant operators of pressurized water reactors (PWR) at the Nuclear Power Training Center Ltd. (NTC) in Japan, training programs were reviewed referring to US training programs. A systematic approach is deployed to them, which mainly consist of on-the-job training and the NTC training courses to meet the needs of all operators from beginners to experienced veterans according to their experiences and objectives. The NTC training is conducted using the simulators that simulate the nuclear power plant dynamics through the use of computers. The operators trained at the NTC work in the central control room of every PWR power plant. The NTC also carries out the qualification examinations for the shift managers. (T. Tanaka)

Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2

International Nuclear Information System (INIS)

Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P.

1993-12-01

The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ''Pressurizer spray valve faulty opening'' presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data

Specific safety aspects of the water-steam cycle important to nuclear power plant project

International Nuclear Information System (INIS)

Lobo, C.G.

1986-01-01

The water-steam cycle in a nuclear power plant is similar to that used in conventional power plants. Some systems and components are required for the safe nuclear power plant operation and therefore are designed according to the safety criteria, rules and regulations applied in nuclear installations. The aim of this report is to present the safety characteristics of the water-steam cycle of a nuclear power plant with pressurized water reactor, as applied for the design of the nuclear power plants Angra 2 and Angra 3. (Author) [pt

Health protection and industrial safety. Nuclear power plants

International Nuclear Information System (INIS)

1987-03-01

The standard applies to components of the primary circuit including its auxiliary facilities, and of the secondary circuit of nuclear power plants with pressurized water reactors; to lifting gear and load take-ups for the transport of nuclear fuel and primary circuit components, and to elevators within the containment. Part 2 specifies testing, test periods, test methods, and documentation

Annotated bibliography of safety-related occurrences in pressurized-water nuclear power plants as reported in 1976

International Nuclear Information System (INIS)

Scott, R.L.; Gallaher, R.B.

1977-01-01

The bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at pressurized-water reactor nuclear power plants in 1976. Included are 1264 abstracts that describe incidents, failures, and design construction deficiencies experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables summarizing the information contained in the bibliography are presented. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). A few of the unique events that occurred during the year are reviewed in detail

Annotated bibliography of safety-related occurrences in pressurized-water nuclear power plants as reported in 1975

International Nuclear Information System (INIS)

Scott, R.L.; Gallaher, R.B.

1976-07-01

The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at pressurized-water reactor nuclear power plants in 1975. The report includes 1097 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables summarizing the information contained in the bibliography are presented. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). A few of the unique events that occurred during the year are reviewed in detail

Annotated bibliography of safety-related occurrences in pressurized-water nuclear power plants as reported in 1976

Energy Technology Data Exchange (ETDEWEB)

Scott, R.L.; Gallaher, R.B.

1977-08-01

The bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at pressurized-water reactor nuclear power plants in 1976. Included are 1264 abstracts that describe incidents, failures, and design construction deficiencies experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables summarizing the information contained in the bibliography are presented. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). A few of the unique events that occurred during the year are reviewed in detail.

Annotated bibliography of safety-related occurrences in pressurized-water nuclear power plants as reported in 1975

Energy Technology Data Exchange (ETDEWEB)

Scott, R.L.; Gallaher, R.B.

1976-07-01

The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at pressurized-water reactor nuclear power plants in 1975. The report includes 1097 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables summarizing the information contained in the bibliography are presented. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). A few of the unique events that occurred during the year are reviewed in detail.

Steam explosions-induced containment failure studies for Swiss nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Zuchuat, O.; Schmocker, U. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland); Esmaili, H.; Khatib-Rahbar, M.

1998-01-01

The assessment of the consequences of both in-vessel and ex-vessel energetic fuel-coolant interaction for Beznau (a Westinghouse pressurized water reactor with a large, dry containment), Goesgen (a Siemens/KWU pressurized water reactor with a large, dry containment) and Leibstadt (a General Electric boiling water reactor-6 with a free standing steel, MARK-III containment) nuclear power plants is presented in this paper. The Conditional Containment Failure Probability of the steel containment of these Swiss nuclear power plants is determined based on different probabilistic approaches. (author)

Deuterium ingress at rolled joints in Embalse nuclear power plant

International Nuclear Information System (INIS)

Ramos Nervi, J. E.; Schroeter, F.

2013-01-01

Deuterium ingress model at the Rolled Joint has been extensively used for CANDU Nuclear Power Plants Operators in the Life Management of the Pressure Tubes. The importance of understanding the model is vital to avoid delayed hydride cracking at the Rolled Joint. This work reports the first step on develop the model presented on literature to be used in Argentinean CANDU 6, Embalse Nuclear Power Plant. (author)

BWR type nuclear power plant

International Nuclear Information System (INIS)

Matsumoto, Kosuke.

1991-01-01

In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

Large nuclear steam turbine plants

International Nuclear Information System (INIS)

Urushidani, Haruo; Moriya, Shin-ichi; Tsuji, Kunio; Fujita, Isao; Ebata, Sakae; Nagai, Yoji.

1986-01-01

The technical development of the large capacity steam turbines for ABWR plants was partially completed, and that in progress is expected to be completed soon. In this report, the outline of those new technologies is described. As the technologies for increasing the capacity and heightening the efficiency, 52 in long blades and moisture separating heaters are explained. Besides, in the large bore butterfly valves developed for making the layout compact, the effect of thermal efficiency rise due to the reduction of pressure loss can be expected. As the new technology on the system side, the simplification of the turbine system and the effect of heightening the thermal efficiency by high pressure and low pressure drain pumping-up method based on the recent improvement of feed water quality are discussed. As for nuclear steam turbines, the actual records of performance of 1100 MW class, the largest output at present, have been obtained, and as a next large capacity machine, the development of a steam turbine of 1300 MWe class for an ABWR plant is in progress. It can be expected that by the introduction of those new technologies, the plants having high economical efficiency are realized. (Kako, I.)

Simulation of a Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant

International Nuclear Information System (INIS)

Reis Martins Junior, L.L. dos.

1980-01-01

The following work intends to perform the digital simulation, of the Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant for control systems design and analysis purposes. There are mathematical models for the reactor, the steam generator, the pressurizer and for transport lags of the coolant in the primary circuit. Nevertheless no one control system has been considered to permit any user the inclusion in the more convenient way of the desired control systems' models. The characteristics of the system in consideration are fundamentally equal to the ones of Almirante Alvaro Alberto Nuclear Power Plant, Unit I (Angra I) obtained in the Final Safety Analysis Report at Comissao Nacional de Energia Nuclear. (author)

Pump selection and application in a pressurized water reactor electric generating plant

International Nuclear Information System (INIS)

Kitch, D.M.

1985-01-01

Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

Investigations of gas explosions in a nuclear coal gasification plant

International Nuclear Information System (INIS)

Schulte, K.

1981-01-01

The safety research program on gas cloud explosions is performed in the context of the German project of the Prototype Plant Nuclear Process Heat. By the work within this project, it is tried to extend the use of nuclear energy to non-electric application. The programme comprises efforts in several scientific disciplines. The final goal is to provide a representative pressure-time-function or a set of such functions. These functions should be the basis for safe design and construction of the nuclear reactor system of a coal gasification plant. No result yet achieved contradicts the assumption that released process gas is only able to deflagrate. It should be possible to demonstrate that, if unfavourable configurations are avoided, a design pressure of 300 mbar is sufficient to withstand an explosion of process gas; this pressure should never be exceeded by process gas explosions irrespective of gas mass released and distance to release point, except possibly in relatively small areas

Linguistic control of a nuclear power plant

International Nuclear Information System (INIS)

Feeley, J.J.; Johnson, J.C.

1987-01-01

A multivariable linguistic controller based on fuzzy set theory is discussed and its application to a pressurized water nuclear power plant control is illustrated by computer simulation. The nonlinear power plant simulation model has nine states, two control inputs, one disturbance input, and two outputs. Although relatively simple, the model captures the essential coupled nonlinear plant dynamics and is convenient to use for control system studies. The use of an adaptive version of the controller is also demonstrated by computer simulation

Safety analysis of Oi nuclear power plant

International Nuclear Information System (INIS)

1979-01-01

The transient phenomena in Oi nuclear power plant were analyzed, especially on the water level fluctuation and the capability of natural circulation in the primary loop, under the assumptions that the feed water for steam generators is totally lost, and the relief valve on the pressurizer, which is actuated due to the pressure rise in the primary system, is stuck and kept open. These assumptions are related to the TMI accident. The analysing conditions are 1) the main feed water flow is totally lost suddenly during the rated power operation of the reactor, 2) two motor-driven auxiliary feed water pumps are started manually fifteen minutes after the accident initiation, 3) one relief valve on the pressurizer is opened fifteen seconds after the accident initiation and kept open, 4) the reactor is scrammed thirty three seconds after the accident initiation, 5) the turbine is tripped 33.5 seconds after the accident initiation, etc. Two cases were analysed, namely 3,800 seconds and 1,200 seconds after the accident initiation. The analytical code RELEP4/Mod5/U2/J1 was utilized for this analysis. The level fluctuation in the pressurizer after the accident initiation, the flow rate fluctuation through the pressurizer relief valve, especially that of steam, liquid single phase and two phase flows, the water level in the upper plenum in the pressure vessel, the change of flow rate at core inlet, the average pressure in the core, and the temperature fluctuation of coolant in the core, the variation of void fraction in the core, and the change of surface temperature of fuel rods are presented as the analysis results, and they are evaluated. It is recognized that the plant safety is kept under the assumed accident conditions in the Oi nuclear power plant. (Nakai, Y.)

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Sea water pumping-up power plant system combined with nuclear power plant

International Nuclear Information System (INIS)

Ichiki, Tadaharu; Tanaka, Masayuki.

1991-01-01

It is difficult to find a site suitable to construction for a sea water pumping-up power plant at a place relatively near the electric power consumption area. Then, a nuclear power plant is set at the sea bottom or the land portion of a sea shore near the power consumption area. A cavity is excavated underground or at the bottom of the sea in the vicinity of the power plant to form a lower pond, and the bottom of the sea, as an upper pond and the lower pond are connected by a water pressure pipe and a water discharge pipe. A pump water turbine is disposed therebetween, to which electric power generator is connected. In addition, an ordinary or emergency cooling facility in the nuclear power plant is constituted such that sea water in the cavity is supplied by a sea water pump. Accordingly, the sea water pumping-up plant system in combination with the nuclear power plant is constituted with no injuring from salts to animals and plants on land in the suburbs of a large city. The cost for facilities for supplying power from a remote power plant to large city areas and power loss are decreased and stable electric power can be supplied. (N.H.)

Current developments in mechanized non-destructive testing in nuclear power plants

International Nuclear Information System (INIS)

Zeilinger, R.

2008-01-01

Nuclear power plants require frequent in-service activities to be carried out conscientiously in areas potentially hazardous to human operators (because of the associated radiation exposure), such as non-destructive testing of pressurized components of the steam system. Locations to be inspected in this way include the reactor pressure vessel, core internals, steam generators, pressurizers, and pipes. The codes to be used as a basis of these inspections demand high absolute positioning and repeating accuracy. These requirements can be met by mechanized test procedures. Accordingly, many new applications of, mostly mobile, robots have been developed over the past few years. The innovative control and sensor systems for stationary and mobile robots now on the market offer a potential for economic application in a large number of new areas in inspection, maintenance and service in nuclear power plants. More progress in this area is expected for the near future. Areva NP founded the new NDT Center, NETEC (Non-destructive Examination Technical Center), as a global technical center for non-destructive materials testing. NETEC is to advance research and development of all basic NDT technologies, robotics included. For many years, intelligeNDT has offered solutions and products for a variety of inspection and testing purposes and locations in nuclear power plants and is involved in continuous further development of the experience collected in nuclear power plants on the spot. (orig.)

PWR heavy equipments manufacture for nuclear power plants

International Nuclear Information System (INIS)

Boury, C.; Terrien, J.F.

1983-10-01

The manufacture of boilers has been imported by the French nuclear program to the societe FRAMATOME. FRAMATOME, because of the size of this market, has constructed two special plants for manufacturing of nuclear components (vapor generators, reactor tanks, pressurizers); these two high technical facilities are presented: production, staff training, technical overseas assistance, and technical and economical repercussions on the industrial vicinity [fr

Psychological characteristics of licensed nuclear power plant operators

International Nuclear Information System (INIS)

Sajwaj, T.; Ford, T.; McGee, R.K.

1987-01-01

The safe production of electricity by nuclear power plants has been the focus of considerable attention. Much of this concern has been focused on equipment and procedural issues, with less attention to the psychological factors that affect the operations staff of the plants, i.e., those individuals who are most directly responsible for a plant's operations. Stress and type A qualities would be significant for these individuals because of their relationships to job performance and health. Of equal significance would be work-related factors, such as job involvement and work pressure. Also of interest would be hostile tendencies because of the need for cooperation and communications among operations staff. Two variables could influence these psychological factors. One is the degree of responsibility for a plant's nuclear reactors. The individuals with the greatest responsibility are licensed by the US Nuclear Regulatory Commission (NRC). There are also individuals with less direct responsibilities who are not licensed. A second variable is the operating status of the plant, whether or not the plant is currently producing electricity. Relative to ensuring the safe operation of nuclear power plants, these data suggest a positive view of licensed operators. Of interest are the greater stress scores in the licensed staff of the operating plant in contrast with their peers in the nonoperating plant

UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

International Nuclear Information System (INIS)

Watson, P.S.

1990-01-01

Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

The safety of nuclear power plants in Eastern Europe

International Nuclear Information System (INIS)

Hoehn, J.; Niehaus, F.

1997-01-01

Nuclear power plant operators and nuclear organizations from the West and from the East cooperate at many levels. The G7 and G24 nations have taken it upon themselves to improve the safety of Eastern nuclear power plants. The European Union has launched support programs, i.e. Technical Assistance to the Commonwealth of Independent States (Tacis) and Pologne-Hangrie: Aide a la Reconstruction Economique (Phare), and founded the European Bank for Reconstruction and Development. The countries of Central and Eastern Europe operate nuclear power plants equipped with VVER-type pressurized water reactors and those equipped with RBMK-type reactors. The safety of these two types of plants is judged very differently. Among the VVER plants, a distinction is made between the older and the more recent 440 MWe lines and the 1000 MWe line. Especially the RBMK plants (Chernobyl-type plants) differ greatly as a function of location and year of construction. Even though they do not meet Western safety standards and at best can be backfitted up to a certain level, it must yet be assumed that they will remain in operation to the end of their projected service lives for economic reasons. (orig.) [de

Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

International Nuclear Information System (INIS)

Ware, A.G.; Morton, D.K.; Nitzel, M.E.

1995-03-01

Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four U.S. nuclear steam supply system vendors. For each facility, six locations were studied, including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This report discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

A model for structural analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A. de.

1987-01-01

Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Chen, Yen-Shu, E-mail: yschen@iner.org.t [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

2011-05-15

Research highlights: The Chinshan Mark I containment pressure-temperature responses are analyzed. GOTHIC is used to calculate the containment responses under three pipe break events. This study is used to support the Chinshan Stretch Power Uprate (SPU) program. The calculated peak pressure and temperature are still below the design values. The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 {sup o}C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 {sup o}C). Additionally, the peak drywell temperature of 155.3 {sup o}C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 {sup o}C, which is below the pool temperature used for evaluating the

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon

2011-01-01

Research highlights: → The Chinshan Mark I containment pressure-temperature responses are analyzed. → GOTHIC is used to calculate the containment responses under three pipe break events. → This study is used to support the Chinshan Stretch Power Uprate (SPU) program. → The calculated peak pressure and temperature are still below the design values. → The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 o C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 o C). Additionally, the peak drywell temperature of 155.3 o C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 o C, which is below the pool temperature used for evaluating the

The function of people in a nuclear power plant in operation

International Nuclear Information System (INIS)

Pickel, E.

1986-01-01

Based mainly on the operating experience of Obrigheim nuclear power plant - a two-loop pressurized water reactor with about 120,000 operating hours - the function of people in the very complex technical field of operation including supervision, maintenance and repair of a nuclear power plant is presented with special attention to safety related activities. The boundaries are indicated where automatic actions for safety and economic reasons are necessary and where manually operated systems are possible or even desirable. Typical examples and figures will show, how the operating people influenced the operation and availability of Obrigheim nuclear power plant. (orig./GL)

Experience with ALARA and ALARA procedures in a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Abrahamse, J.C. [N.V. Elektriciteits-Produktiemaatschappij Zuid-Nederland (Netherlands)

1995-03-01

The nuclear power plant Borssele is a Siemens two-loop Pressurized Water Reactor having a capacity of 480 MWe and in operation since 1973. The nuclear power plant Borssle is located in the southwest of the Netherlands, near the Westerschelde River. In the first nine years of operation the radiation level in the primary system increased, reaching a maximum in 1983. The most important reason for this high radiation level was the cobalt content of the grid assemblies of the fuel elements. After resolving this problem, the radiation level decreased to a level comparable with that of other nuclear power plants.

The 1978 first in-service inspection of the reactor pressure vessel of the second unit of the Greifswald nuclear power plant

International Nuclear Information System (INIS)

Pastor, D.; Busch, R.; Hildebrandt, E.; Redlich, K.H.

1979-01-01

The reactor pressure vessel and the primary coolant circuit of the second 440-MW(e) unit of the Greifswald nuclear power plant were subjected to an in-service inspection. Extent of the inspection, development and construction of a reactor inspection container as well as the nondestructive materials testing methods used are described. Further, problems of performing the inspection, such as needs of time and personnel and radiation exposure, are considered. Finally, it is stated that the reactor pressure vessel was in safe operating state. (author)

Protecting nuclear power plants. Chapter 2. On the importance of the security and safety of the reactor pressure vessel to external threats

International Nuclear Information System (INIS)

Ballesteros, A.; Gonzalez, J.; Debarberis, L.

2006-01-01

Nuclear power plants have blong been recognized as potential targets of terrorist attacks, and critics have long questioned the adequacy of the existing measures to defend against such attacks. The 11-S 2001, 11-M 2004 and 7-J 2005 attacks in USA, Spain and UK illustrated the deadly intention and abilities of modern terrorist groups. These attacks also brought to surface long standing concerns about the vulnerability of nuclear installations to possible terrorist attacks. Commercial nuclear reactors contain large inventory of radioactive fission products which, if dispersed, could pose a direct radiation hazard on the population. The reactor pressure vessel (RPV), which contains the nuclear fuel, is the most critical component of the plant. This paper shows that small amount of explosive material can produce irreversible damage in the RPV and the release of radioactive material. Therefor, access of working personal to the vicinity of the RPV during the refuelling outage should be stricktly limited. It should be considered a high priority security issue

Evaluation of the ultimate pressure capacity of rectangular HVAC ducts for nuclear pwoer plants

International Nuclear Information System (INIS)

Wedellsborg, B.W.

1984-01-01

Typical Category I HVAC ducts in a nuclear plant must be designed for loads and load combinations including positive and negative pressure loads which are generated due to the normal operation and postulated accident conditions. These pressure loads most often govern the design of the HVAC ducts. Structural design criteria are presently based on the AISI Code which limits the duct panel width-to-thickness ratio to a maximum of 500 and the maximum height-to-thickness ratio to 200, unless it can be shown by structural tests that larger ratios can be used. Test Programs performed on rectangular HVAC ducts subjected to vacumm loads have substantiated the use of ducts having panel width to thickness ratios of up to 1600. The results of the test programs were subsequently incorporated into the design through a more rational analytical design method which was developed from and correlates well with the test results. The purpose of this paper is to present the analytical design method and its correlation with the test results. Simple formulae for the design of rectangular HVAC ducts are presented. Lower bound values of duct sheet, and stiffener ultimate loads are derived, and correlated with recent test results. Analytically predicted ultimate pressures are also compared with other available duct test data

Nuclear power plant

International Nuclear Information System (INIS)

Schabert, H.P.

1976-01-01

A nuclear power plant is described which includes a steam generator supplied via an input inlet with feedwater heated by reactor coolant to generate steam, the steam being conducted to a steam engine having a high pressure stage to which the steam is supplied, and which exhausts the steam through a reheater to a low pressure stage. The reheater is a heat exchanger requiring a supply of hot fluid. To avoid the extra load that would be placed on the steam generator by using a portion of its steam output as such heating fluid, a portion of the water in the steam generator is removed and passed through the reheater, this water having received at least adequate heating in the steam generator to make the reheater effective, but not at the time of its removal being in a boiling condition

Periodic inspection of CANDU nuclear power plant containment components

International Nuclear Information System (INIS)

1989-09-01

This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

Temporal neural network for the identification of nuclear power plant transients

International Nuclear Information System (INIS)

Uluyol, O.; Ragheb, M.

1993-01-01

In this paper a layered spatiotemporal neural network is proposed for the identification of nuclear power plant transients. The developed layered spatiotemporal network is inspired by the formal avalanche structure developed by S. Grossberg and offers advantages compared with the stationary pattern approach using the perceptron paradigm. Each layer in the network is trained to recognize a separate time-dependent accident scenario. Within each scenario, the temporal behavior of the relevant parameters such as pressurizer pressure, pressurizer water volume, cold and hot legs temperatures, vessel flow, and power, are considered. Numerical cases are considered where the proposed methodology is applied to two nuclear power plant anticipated transient scenarios: the Station Blackout and the Anticipated Transient without Scram transients in a pressurized water reactor . The transient signatures used were generated by modeling the accidents using RELAP5/MOD2, a best-estimate thermal-hydraulics numerical code. The ability of the proposed layered spatiotemporal network to operate at different noise levels is investigated. Its incorporation within an Insightful Algorithm and Anticipatory Systems context for identifying and in predicting the course of nuclear transients is discussed

Feasibility study of a dedicate nuclear desalination system: Low-pressure inherent heat sink nuclear desalination plant (LIND)

Energy Technology Data Exchange (ETDEWEB)

Kim, Ho Sik; No, Hee Cheon; Jo, Yu Gwan; Wivisono, Andhika Feri; Park, Byung Ha; Choi, Jin Young; Lee, Jeong Ik; Jeong, Yong Hoon; Cho, Nam Zin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-04-15

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MW{sub th} and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

Feasibility study of a dedicated nuclear desalination system: Low-pressure Inherent heat sink Nuclear Desalination plant (LIND

Directory of Open Access Journals (Sweden)

Ho Sik Kim

2015-04-01

Full Text Available In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal–hydraulic and neutronic design requirements. In a thermal–hydraulic analysis using an analytical method based on the Wooton–Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MWth and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

Nuclear power plant operating experience. Annual report, 1978

International Nuclear Information System (INIS)

Beebe, M.R.

1979-12-01

This report is the fifth in a series of reports issued annually that summarizes the operating experience of US nuclear power plants in commercial operation. Power generation statistics, plant outages, reportable occurrences, fuel element performance, occupational radiation exposure for each plant are presented. Summary highlights of these areas are discussed. The report includes 1978 data from 65 plants - 25 boiling water reactor plants and 40 pressurized water reactor plants. Discussion of radioactive effluents which has been a part of this report in previous years, has not been included in this issue because of late acquisition of data

PCB transformer fires: the risk in nuclear power plants

International Nuclear Information System (INIS)

Blackmon, K.

1988-01-01

It is estimated that 1/2 of the present nuclear power plants operate with PCB-filled transformer equipment. In an attempt to obtain better estimates of clean-up costs in a nuclear power plant under reasonable-loss scenarios, a study was commissioned. This study was a joint venture between Blackmon-Mooring Steamatic Technologies, Inc., (BMS-TECH) and M and M Protection Consultants. This joint study was conducted at a typical pressurized-water reactor plant consisting of two 1000-MW units. Three specific scenarios were selected and analyzed for this typical power plant. These scenarios were: (1) an electrical failure of a transformer in an isolated switch gear room; (2) a transformer exposed to a 55-gallon transient combustion oil fire in the auxiliary building; and (3) a PCB transformer involved in a major turbine lube fire in the turbine building. Based on results of this study, the insurance carriers for this industry implemented an adjustment in their rate structures for nuclear power plants that have PCB equipment

Plant for the delivery of long-distance steam combined with a nuclear power plant

International Nuclear Information System (INIS)

Schueller, K.H.

1976-01-01

It is proposed that long-distance steam should not be directly discharged in order to avoid each posibility of spreading radioactively contaminated steam. As a heat transmitter, a surface heat exchanger should be chosen, the heating steam of the nuclear power station heating pressurized water whose pressure is higher then that of the heating steam. Long-distance steam generation then results from expanding the pressurized water. The plant is described in detail. (UWI) [de

Configuration management of plant modifications for nuclear power plants

International Nuclear Information System (INIS)

Ritsch, W.J.

1987-01-01

Due to the increasing complexity of nuclear power plant operation, regulatory pressure, and the large numbers of people required to operate and support the stations, the control of plant modifications at these plants needs to be expanded and improved. The aerospace and defense industries, as well as the owners or operators of large energy projects have established configuration management programs (CMPs) to control plant design changes. These programs are composed of well-defined functions for identifying, evaluating, recording, tracking, issuing, and documenting the established baseline conditions, as well as required changes to these baseline conditions. The purpose of this paper is to describe a recommended CMP for plant modifications consisting of a computerized data base installed on the utility's computer to provide a central storage of plant design and operations data necessary to control the following activities as they are affected by plant design changes: training; record management; operations; maintenance; health physics; planning/scheduling; procurement/inventory control; outage management (including modifications); and emergency response

Fukushima Nuclear Accident, the Third International Severe Nuclear Power Plant Accident

International Nuclear Information System (INIS)

Rashad, S.M.

2013-01-01

Japan is the world's third largest power user. Japan's last remaining nuclear reactor shutdown on Saturday 4 Th of May 2012 leaving the country entirely nuclear free. All of 50 of the nation's operable reactors (not counting for the four crippled reactors at Fukushima) are now offline. Before last year's Fukushima nuclear disaster, the country obtained 30% of its energy from nuclear plants, and had planned to produce up to 50% of its power from nuclear sources by 2030. Japan declared states of emergency for five nuclear reactors at two power plants after the units lost cooling ability in the aftermath of Friday 11 March 2011 powerful earthquake. Thousands of (14000) residents were immediately evacuated as workers struggled to get the reactors under control to prevent meltdowns. On March 11 Th, 2011, Japan experienced a sever earthquake resulting in the shutdown of multiple reactors. At Fukushima Daiichi site, the earthquake caused the loss of normal Ac power. In addition it appeals that the ensuing tsunami caused the loss of emergency Ac power at the site. Subsequent events caused damage to fuel and radiological releases offsite. The spent fuel problem is a wild card in the potentially catastrophic failure of Fukushima power plant. Since the Friday's 9.0 earthquake, the plant has been wracked by repeated explosions in three different reactors. Nuclear experts emphasized there are significant differences between the unfolding nuclear crisis at Fukushima and the events leading up to the Chernobyl disaster in 1986. The Chernobyl reactor exploded during a power surge while it was in operation and released a major cloud of radiation because the reactor had no containment structure around to. At Fukushima, each reactor has shutdown and is inside a 20 cm-thick steel pressure vessel that is designed to contain a meltdown. The pressure vessels themselves are surrounded by steel-lined, reinforced concrete shells. Chernobyl disaster was classified 7 on the International

Waste generation comparison: Coal-fired versus nuclear power plants

International Nuclear Information System (INIS)

LaGuardia, T.S.

1998-01-01

Low-level radioactive waste generation and disposal attract a great deal of attention whenever the nuclear industry is scrutinized by concerned parties, be it the media, the public, or political interests. It is therefore important to the nuclear industry that this issue be put into perspective relative to other current forms of energy production. Most of the country's fossil-fueled power comes from coal-fired plants, with oil and gas as other fuel sources. Most of the generated waste also comes from coal plants. This paper, therefore, compares waste quantities generated by a typical (1150-MW(electric)) pressurized water reactor (PWR) to that of a comparably sized coal-fired power plant

The impact of team building and leadership development on nuclear plant performance

International Nuclear Information System (INIS)

Fiedler, P.B.; Long, R.L.; Childress, J.R.

1988-01-01

Within the nuclear utility industry, the pressures of complex technologies, increasing regulations, and critical public scrutiny create a working environment filled with numerous pressures. The difficult nature of the industry puts a premium on effective teamwork, interdepartmental cooperation, and communication skills. A well-conceived and implemented team building and leadership development program can substantially improve the operating performance of a nuclear plant. This paper describes one such implementation effort at GPU Nuclear Corporation and at the Oyster Creek nuclear generating station (OCNGS) over an 18-month period

Trial application of guidelines for nuclear plant response to an earthquake

International Nuclear Information System (INIS)

Schmidt, W.; Oliver, R.; O'Connor, W.

1993-09-01

Guidelines have been developed to assist nuclear plant personnel in the preparation of earthquake response procedures for nuclear power plants. These guidelines are published in EPRI report NP-6695, ''Guidelines for Nuclear Plant Response to an Earthquake,'' dated December 1989. This report includes two sets of nuclear plant procedures which were prepared to implement the guidelines of EPRI report NP-6695. The first set were developed by the Toledo Edison Company Davis-Besse plant. Davis-Besse is a pressurized water reactor (PWR) and contains relatively standard seismic monitoring instrumentation typical of many domestic nuclear plants. The second set of procedures were prepared by Yankee Atomic Electric Company for the Vermont Yankee facility. This plant is a boiling water reactor (BWR) with state-of-the-art seismic monitoring and PC-based data processing equipment, software developed specifically to implement the OBE Exceedance Criterion presented in EPRI report NP-5930, ''A Criterion for Determining Exceedance of the operating Basis Earthquake.'' The two sets of procedures are intended to demonstrate how two different nuclear utilities have interpreted and applied the EPRI guidance given in report NP-6695

Health protection and industrial safety. Nuclear power plants

International Nuclear Information System (INIS)

1987-03-01

The standard applies to components of the primary circuit including its auxiliary facilities, and of the secondary circuit of nuclear power plants with pressurized water reactors; to lifting gear and load take-ups for the transport of nuclear fuel and primary circuit components; to elevators within the containment, electrical installations, and piping and valves of radiation protection monitoring equipment. Part 1 defines the terms and specifies engineered safety requirements

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Staunton, R.H.; Cox, D.F. [Oak Ridge National Lab., TN (United States)

1995-03-01

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail.

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

International Nuclear Information System (INIS)

Staunton, R.H.; Cox, D.F.

1995-03-01

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail

SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts

International Nuclear Information System (INIS)

Duffey, R.B.; Pioro, I.L.; Gabaraev, B.A.; Kuznetsov, Yu. N.

2006-01-01

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

Nuclear power plant control room task analysis. Pilot study for pressurized water reactors

International Nuclear Information System (INIS)

Barks, D.B.; Kozinsky, E.J.; Eckel, S.

1982-05-01

The purposes of this nuclear plant task analysis pilot study: to demonstrate the use of task analysis techniques on selected abnormal or emergency operation events in a nuclear power plant; to evaluate the use of simulator data obtained from an automated Performance Measurement System to supplement and validate data obtained by traditional task analysis methods; and to demonstrate sample applications of task analysis data to address questions pertinent to nuclear power plant operational safety: control room layout, staffing and training requirements, operating procedures, interpersonal communications, and job performance aids. Five data sources were investigated to provide information for a task analysis. These sources were (1) written operating procedures (event-based); (2) interviews with subject matter experts (the control room operators); (3) videotapes of the control room operators (senior reactor operators and reactor operators) while responding to each event in a simulator; (4) walk-/talk-throughs conducted by control room operators for each event; and (5) simulator data from the PMS

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

2007-03-21

In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

French nuclear plant safeguard pump qualification testing: EPEC test loop

International Nuclear Information System (INIS)

Guesnon, H.

1985-01-01

This paper reviews the specifications to which nuclear power plant safeguard pumps must be qualified, and surveys the qualification methods and program used in France to verify operability of the pump assembly and major pump components. The EPEC test loop is described along with loop capabilities and acheivements up to now. This paper shows, through an example, the Medium Pressure Safety Injection Pump designed for service in 1300 MW nuclear power plants, and the interesting possibilities offered by qualification testing

Applications of computer based safety systems in Korea nuclear power plants

International Nuclear Information System (INIS)

Won Young Yun

1998-01-01

With the progress of computer technology, the applications of computer based safety systems in Korea nuclear power plants have increased rapidly in recent decades. The main purpose of this movement is to take advantage of modern computer technology so as to improve the operability and maintainability of the plants. However, in fact there have been a lot of controversies on computer based systems' safety between the regulatory body and nuclear utility in Korea. The Korea Institute of Nuclear Safety (KINS), technical support organization for nuclear plant licensing, is currently confronted with the pressure to set up well defined domestic regulatory requirements from this aspect. This paper presents the current status and the regulatory activities related to the applications of computer based safety systems in Korea. (author)

The analysis of cracks in high-pressure piping and their effects on strength and lifetime of construction components at the Ignalina nuclear plant

Energy Technology Data Exchange (ETDEWEB)

Aleev, A.; Petkevicius, K.; Senkus, V. [and others

1997-04-01

A number of cracks and damages of other sorts have been identified in the high-pressure parts at the Ignalina Nuclear Plant. They are caused by inadequate production- and repair technologies, as well as by thermal, chemical and mechanical processes of their performance. Several techniques are available as predictions of cracks and other defects of pressurized vessels. The choice of an experimental technique should be based on the level of its agreement with the actual processes.

The main features of control and operation of steam turbines at nuclear power plants

International Nuclear Information System (INIS)

Czinkoczky, B.

1981-01-01

The output and speed control of steam turbines at nuclear power plants as well as the combination of both controls are reviewed and evaluated. At the same time the tasks of unit control at nuclear power plants, the control of steady main steam pressure and medium pressure of primary circuit, further the connection of reactor and turbine controls and the self-controlling properties of pressurized water reactor are dealt with. Hydraulic and electro-hydraulic speed control, the connection of cach-up dampers and speed control and the application of electro-hydraulic signal converters are discussed. The accomplishment of protection is also described. (author)

Results of safety analysis on PWR type nuclear power plants with two and three loops

International Nuclear Information System (INIS)

1979-01-01

The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

Pipe restraints for nuclear power plants

International Nuclear Information System (INIS)

Keever, R.E.; Broman, R.; Shevekov, S.

1976-01-01

A pipe restraint for nuclear power plants in which a support member is anchored on supporting surface is described. Formed in the support member is a semicylindrical wall. Seated on the semicylindrical wall is a ring-shaped pipe restrainer that has an inner cylindrical wall. The inner cylindrical wall of the pipe restrainer encircles the pressurized pipe. In a modification of the pipe restraint, an arched-shaped pipe restrainer is disposed to overlie a pressurized pipe. The ends of the arch-shaped pipe restrainer are fixed to support members, which are anchored in concrete or to a supporting surface. A strap depends from the arch-shaped pipe restrainer. The pressurized pipe is supported by the depending strap

Data feature: World nuclear power plant capacity 1991

International Nuclear Information System (INIS)

Anon.

1992-01-01

At this point, the future of the nuclear power industry remains largely in doubt. The gloomy predictions about global warming have done little to convince politicians and the public of the benefits of nuclear power. Meanwhile, the setbacks to nuclear have continued apace: The United States has failed to take the expected lead in ordering new nuclear plants. And President-elect Bill Clinton does not consider nuclear a major part of his energy strategy. The situation looks equally bleak in other countries. Canada's biggest utility, Ontario Hydro, was forced under intense political pressure to defer its ambitious nuclear expansion program until after the year 2010. In Europe, the suspension of France's Superphenix fast-breeder reactor in June could stop progress on the technology indefinitely. And the Finnish parliament dropped plans for expansion of nuclear power from its national energy strategy. Developing and semi-industrialized countries, such as Brazil and Argentina, have shown little progress, taking upwards of twenty years to complete plants already under construction. Nuclear's problems seem always to hinge on economics. Nuclear has little chance of revival during the current global recession, especially in countries fighting for their long-term economic survival. That is why NUKEM believes nuclear power will not grow much in the CIS and Eastern Europe beyond the projects already in the advanced stages of construction. What's more, the longer countries such as Italy, the Netherlands, Spain, Switzerland and Finland keep their nuclear expansion plans on hold, the harder it will be to get the political support to restart them. So far in 1992, only two nuclear plants, with a combined capacity of 1,520 MWe, have gone into commercial operation. One more 1,330 MWe reactor may start up by year's end. By then, NUKEM expects world nuclear plant capacity to stand at 330.3 GWe

An experimental study of assessment of weld quality on fatigue reliability analysis of a nuclear pressure vessel

International Nuclear Information System (INIS)

Dai Shuhe

1993-01-01

The steam generator in PWR primary coolant system China of Qinshan Nuclear Power Plant is a crucial unit belonging to the category of nuclear pressure vessel. The purpose of this research work is to make an examination of the weld quality of the steam generator under fatigue loading and to assess its reliability by using the experimental results of fatigue test of material of nuclear pressure vessel S-271 (Chinese Standard) and of qualified tests of welded seams of a simulated prototype of bottom closure head of the steam generator. A guarantee of weld quality is proposed as a subsequent verification for China National Nuclear Safety Supervision Bureau. The results of reliability analysis reported in this work can be taken as a supplementary material of Probabilistic Safety Assessment (PSA) of Qinshan Nuclear Power Plant. According to the requirement of Provision II-1500 cyclic testing, ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, a simulated prototype of the bottom closure head of the steam generator was made for qualified tests. To find the quantified results of reliability assessment by using the testing data, two proposals are presented

In-service diagnostic systems of steam generators, pressurizers and other components of WWER type nuclear power plants

International Nuclear Information System (INIS)

Matal, O.

1988-01-01

A detailed description is presented of the systems of vibration inspections and systems of determining residual service life, implemented as in-service diagnostic systems for steam generators and pressurizers at the Dukovany nuclear power plant. Low temperature accelerometers of the KD or KS type and high temperature accelerometers CA 91 are used as vibration sensors. In the system of vibration inspection a total of 64 vibration measuring chains of Czechoslovak make and design are installed in the power plant. Systems are being built for determining residual service life which consist of 75 special chains for heat monitoring with thermocouples installed on selected assemblies of the steam generators and the pressurizers serving to monitor and evaluate heat stress. Also included in the system for determining residual service life are 16 routes for water withdrawal from steam generators. Their purpose is to make in-service determinations of places of biggest concentrations of impurities in secondary water, to determine the biggest local chemical exposure of primary collector and heat exchange tube materials and to optimize the size and place of leachate withdrawal. (Z.M.). 2 figs., 2 tabs., 15 refs

Maintenance of process instrumentation in nuclear power plants

CERN Document Server

Hashemian, H M

2006-01-01

Compiles 30 years of practical knowledge gained by the author and his staff in testing the I and C systems of nuclear power plants around the world. This book focuses on process temperature and pressure sensors and the verification of these sensors' calibration and response time.

Chemical process measurements in PWR-type nuclear power plants

International Nuclear Information System (INIS)

Glaeser, E.

1978-01-01

In order to achieve high levels of availability of nuclear power plants equipped with pressurized water reactors, strict standards have to be applied to the purity of coolant and of other media. Chemical process measurements can meet these requirements only if programmes are established giving maximum information with minimum expenditure and if these programmes are realized with effective analytical methods. Analysis programmes known from literature are proved for their usefulness, and hints are given for establishing rational programmes. Analytical techniques are compared with each other taking into consideration both methods which have already been introduced into nuclear power plant practice and methods not yet generally used in practice, such as atomic absorption spectrophotometry, gas chromatography, etc. Finally, based on the state of the art of chemical process measurements in nuclear power plants, the trends of future development are pointed out. (author)

Careful determination of inservice inspection of piping by computer analysis in nuclear power plant

International Nuclear Information System (INIS)

Lim, H. T.; Lee, S. L.; Lee, J. P.; Kim, B. C.

1992-01-01

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in order to predict possibility of crack generation due to thermal stratification phenomena in pipes connected to reactor coolant system of Nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.

Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1981-01-01

Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

Low speed turbines for nuclear power plants

International Nuclear Information System (INIS)

Ugol'nikov, V.V.; Kosyak, Yu.F.; Virchenko, M.A.

1975-01-01

Work of the Kharkov turbine plant on planning and manufacture for nuclear power plants of low-speed (1500 rpm) turbines with a power of 500-1000 MW is described. The selection of a construction diagram for the turbine assembly, determined basically by the presence or absence of parts of average pressure, is considered. Special construction features of the condenser and turbine are described. Turbine K-500, with a rate of 1500 rpm, was calculated for operation in a two-loop nuclear power plant with saturated steam with intermediate separation and two-stage steam regeneration. On the base of this turbine, three models of 1000-MW turbines were developed. The first model has a cylinder of average pressure (TsSD) and a lateral condenser. The second has no TsSD but a low location of the condensers. The third has no TsSD and the condensers are located laterally. Calculations of the heat efficiency of the three types of turbines showed that several advantages are offered by the model with a TsSD. Better aerodynamic properties of the exhaust nozzles and condensers with lateral location allows decreasing the specific heat consumption to 0.5-1% or, at the same power, a 10-20% decrease in cooling water consumption

Nuclear power. Volume 1. Nuclear power plant design

International Nuclear Information System (INIS)

Pedersen, E.S.

1978-01-01

NUCLEAR POWER PLANT DESIGN is intended to be used as a working reference book for management, engineers and designers, and as a graduate-level text for engineering students. The book is designed to combine theory with practical nuclear power engineering and design experience, and to give the reader an up-to-date view of the status of nuclear power and a basic understanding of how nuclear power plants function. Volume 1 contains the following chapters; (1) nuclear reactor theory; (2) nuclear reactor design; (3) types of nuclear power plants; (4) licensing requirements; (5) shielding and personnel exposure; (6) containment and structural design; (7) main steam and turbine cycles; (8) plant electrical system; (9) plant instrumentation and control systems; (10) radioactive waste disposal (waste management) and (11) conclusion

Management of refuelling, modifications and accidental shut-down of nuclear power plant

International Nuclear Information System (INIS)

1996-01-01

This document is the appendix of HAF 0300 (91) 'Code on the Safety of Nuclear Power Plant Operation', which was promulgated by the National Nuclear Safety Administration (NNSA) on March 2, 1994, and has the same legal effect. This appendix is applicable to establish the administrative management procedures for refuelling, modifications and accidental shut-down in the period of operation of pressurized water thermal neutron reactor of nuclear power plants. The NNSA shall be responsible for interpretation of this document

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

International Nuclear Information System (INIS)

2005-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and ware out of components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of heavy water moderated reactors (HWRs), boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

Neutronic calculations for the reactor pressure vessel of Atucha I nuclear power plant

International Nuclear Information System (INIS)

Lerner, Ana M.; Madariaga, Marcelo R.

1999-01-01

In 1974 a surveillance program for the Atucha I nuclear power plant pressure vessel was initiated which included the construction of different types of specimens, distributed in 30 irradiation capsules located under the core at the lower part of some of the fuel channels. The capsules containing the irradiated specimens were withdrawn in two stages; the first set (SET 1) of 15 specimens in 1980 and the second one (SET 2) of the remaining 15, in 1987. Both fracture mechanic tests and dosimetry analysis were carried out by the designer (KWU) for SET1 and by the owner National Atomic Energy Commission (CNEA) for SET2. The calculations performed in the case of SET1 showed that there was a significant spectrum difference between the position where the specimens had been and the reactor pressure vessel (RPV) - inner surface (IS). It was established that the ratio of thermal flux (E 1 MeV) varied, approximately, from 1000 to 10 from the irradiation position to the RPV- IS. The purpose of this report is to show the calculations recently performed at the Nuclear Regulatory Authority, with particular emphasis on the difference in the results generated by the modification to sightly enriched fuel. A simplified 1-D calculations show that there is a slight increase (4% approximately) in the flux along the whole energy range. As it has already been mentioned, this is due, more than to the isotopic composition of the new fuel, to the difference in power density spatial distribution, which is a consequence of a different fuel management, necessary to preserve operational limits below their maximum allowed values with the same total thermal power generated. More detailed calculations are nevertheless foreseen in order to verify these first results. (author)

Rationalizing of construction engineering of nuclear power plants with pressurized water reactors

International Nuclear Information System (INIS)

Schmidt, S.

1977-01-01

Construction of large power plants requires further reduction of construction efforts and the construction period. A new constructional and technological solution has been developed with the steel-cell composite structure applied in the Greifswald nuclear power plant 'Bruno Leuschner' for the first time. Principles of design, fabrication, transport, and mounting are described. The benefits of the method are indicated. (author)

Analysis of liquid relief valves opening demand during pressure increase abnormal scenarios at Embalse nuclear power plant

International Nuclear Information System (INIS)

Bedrossian, Gustavo C.; Gersberg, Sara

2000-01-01

Two hypothetical scenarios have been analyzed where, after an initiating event, Embalse nuclear power plant primary heat transport system could undergo a pressure increase. These abnormal events are a loss of feedwater to the steam generators and a loss of Class IV power supply with Class III restoration. This analysis focuses on primary system liquid relief valves action, specially on their opening demand. Calculation results show that even when these valves are expected to open during the transient, primary system maximum allowable pressure would not be exceeded if they failed to open. System response was also studied in case that one of these relief valves did not close once primary system pressure decreases. For the scenario of loss of feedwater to steam generators, if the degasser-condenser could not be bottled-up, Emergency Cooling Injection conditions would be reached due to a continuos loss of coolant. In case of loss of Class IV -and assuming degasser-condenser bottling-up as service water would not be available- it was observed that primary system should remain pressurized, and with core cooled by thermo siphoning mechanism. (author)

Position control of a floating nuclear power plant

International Nuclear Information System (INIS)

Motohashi, K.; Hamamoto, T.; Sasaki, R.; Kojima, M.

1993-01-01

hydrodynamic pressure acting on the floating plant by a linear potential flow theory. The hydrodynamic pressure is estimated as the superposition of each contribution of incident, scattering and radiation waves. The equations of motion are derived for surge, heave and pitch of the floating plant, taking into account fluid-structure interaction. The response quantities of the floating plant and the tether forces are calculated in the frequency domain by a stationary random vibration theory. Based on the numerical results, the variations in structural and tether responses of the floating plant due to position control are discussed. Furthermore, the resulting response quantities are compared with performance requirements of nuclear power plants

Concept of underground nuclear power plant

International Nuclear Information System (INIS)

Onishi, Sotoaki

1976-01-01

The concept of constructing nuclear power plants on the sea or underground as the future sitting is based on moving the present power plants on the ground with actual results to the sea or underground without changing the design. The underground nuclear power plants have many similar points to underground hydro-electric power stations with many achievements in the construction viewpoint, though they have their proper difficult problems. Of course, it requires to excavate larger underground caves than the case of underground hydro-electric power plants. The maximum dimensions of the caves have been determined through experience in practice. Therefore the developments of design theory and construction technique are desirable in this field. In discussing underground construction, two evaluating methods are considered for the shielding effect of base rocks. The minimum vertical distance up to ground surface from the center of the cave differs depending upon the in-cave pressure, and the conditions of base rock, soil and underground water in case of the accident assumed in the design, and is approximately 60m, if the cave is assumed to be an indefinite cylindrical shape, by the safer side calculation in the above two evaluations. (Wakatsuki, Y.)

Correlation of anti-stress ability of nuclear power plant operators with physiological and psychological indexes

International Nuclear Information System (INIS)

Liao Haihong; Liu Yulong; Li Yuan; Dai Tingting; Qiu Mengyue; Bian Huahui; Chen Weibo; Liu Chunfeng

2012-01-01

Objective: To explore the correlation of the anti-stress ability of nuclear power plant operators with their heart rate, blood pressure and mental health. Methods: A total of 136 male nuclear power plant operators received the continuous performance test (CPT) and mental health test (MHT), while 37 male senior students were used as the control group. Heart rate and blood pressure were recorded before and after the test. The nuclear power plant operators were divided to high score group and low score group by CPT scores. The correlation of the anti-stress ability and blood pressure,heart rate and their mental health was analyzed. Results: The scores of CPT and MHT test with the operators were significantly higher than those of the control group (t=-1.25, 6.25, P<0.05). After the test, the heart rate, systolic pressure and diastolic pressure of the operators were all significantly lower than those of the control group (t=-2.07, 3.28, 2.74, P<0.05). The CPT score of the high score group was lower in the heart rate, systolic pressure and diastolic pressure than the low score group (t=-3.15, -2.78, -2.54, P<0.05), while the scores of CPT and MHT were better (t=0.63, 6.90, P<0.05). CPT scores were negatively correlated with the increasing range of the heart rate, systolic pressure and diastolic pressure (r=- 0.69, -0.94, and -0.62, P<0.05), and positively correlated with the MHT scores (r=0.54, P<0.05). Conclusions: To a certain extent, CPT and MHT test can reflect the anti-stress ability and be used as one of the bases to select nuclear power plant operators and to assess their competence. (authors)

Simulation research on multivariable fuzzy model predictive control of nuclear power plant

International Nuclear Information System (INIS)

Su Jie

2012-01-01

To improve the dynamic control capabilities of the nuclear power plant, the algorithm of the multivariable nonlinear predictive control based on the fuzzy model was applied in the main parameters control of the nuclear power plant, including control structure and the design of controller in the base of expounding the math model of the turbine and the once-through steam generator. The simulation results show that the respond of the change of the gas turbine speed and the steam pressure under the algorithm of multivariable fuzzy model predictive control is faster than that under the PID control algorithm, and the output value of the gas turbine speed and the steam pressure under the PID control algorithm is 3%-5% more than that under the algorithm of multi-variable fuzzy model predictive control. So it shows that the algorithm of multi-variable fuzzy model predictive control can control the output of the main parameters of the nuclear power plant well and get better control effect. (author)

Careful Determination of Inservice Inspection of piping by Computer Analysis in Nuclear Power Plant

International Nuclear Information System (INIS)

Lim, H. T.; Lee, S. L.; Lee, J. P.; Kim, B. C.

1992-01-01

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants

Study on the selection method of feed water heater safety valves in nuclear power plants

International Nuclear Information System (INIS)

Shi Jianzhong; Huang Chao; Hu Youqing

2014-01-01

The selection of the high pressure feedwater heater's safety valve usually follows the principle recommended by HEI standards in thermal power plant. However, the nuclear power plant's heaters generally need to accept a lots of drain from a moisture separator reheater (MSR). When the drain regulating valve was failure in fully open position, a large number of high pressure steam will directly goes into the heater. It make high-pressure heater have a risk of overpressure. Therefore, the safety valve selection of the heaters for nuclear power plants not only need to follow the HEI standards, but also need to check his capacity in certain special conditions. The paper established a calculation method to determine the static running point of the heaters based on characteristic equations of the feed water heater, drain regulating valve and steam extraction pipings, and energy balance principle. The method can be used to calculate the equilibrium pressure of various special running conditions, so further determine whether the capacity of the safety valve meets the requirements of safety and emissions. The method proposed in this paper not only can be used for nuclear power plants, can also be used for thermal power plants. (authors)

Nuclear power plant component protection

International Nuclear Information System (INIS)

Michel, E.; Ruf, R.; Dorner, H.

1976-01-01

Described is a nuclear power plant installation which includes a concrete biological shield forming a pit in which a reactor pressure vessel is positioned. A steam generator on the outside of the shield is connected with the pressure vessel via coolant pipe lines which extend through the shield, the coolant circulation being provided by a coolant pump which is also on the outside of the shield. To protect these components on the outside of the shield and which are of mainly or substantially cylindrical shape, semicylindrical concrete segments are interfitted around them to form complete outer cylinders which are retained against outward separation radially from the components, by rings of high tensile steel which may be interspaced so closely that they provide, in effect, an outer steel cylinder. The invention is particularly applicable to pressurized-water coolant reactor installations

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

Science.gov (United States)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

Nuclear plant simulation using the Nuclear Plant Analyzer

International Nuclear Information System (INIS)

Beelman, R.J.; Laats, E.T.; Wagner, R.J.

1984-01-01

The Nuclear Plant Analyzer (NPA), a state-of-the-art computerized safety analysis and engineering tool, was employed to simulate nuclear plant response to an abnormal transient during a training exercise at the US Nuclear Regulatory Commission (USNRC) in Washington, DC. Information relative to plant status was taken from a computer animated color graphics display depicting the course of the transient and was transmitted to the NRC Operations Center in a manner identical to that employed during an actual event. Recommendations from the Operations Center were implemented during on-line, interactive execution of the RELAP5 reactor systems code through the NPA allowing a degree of flexibility in training exercises not realized previously. When the debriefing was conducted, the RELAP5 calculations were replayed by way of the color graphics display, adding a new dimension to the debriefing and greatly enhancing the critique of the exercise

Atucha I nuclear power plant surveillance programme

International Nuclear Information System (INIS)

Jinchuk, D.

1993-01-01

After a review of the main characteristics of the Atucha I nuclear power plant and its pressure vessel, the embrittlement surveillance capsules and the irradiation conditions are described; Charpy impact tests and tensile tests were performed on the irradiated samples, and results are discussed and compared to theoretical calculations: transition temperature shifts, displacement per atom values. 6 refs., 16 figs., 7 tabs

Atucha I nuclear power plant surveillance programme

Energy Technology Data Exchange (ETDEWEB)

Jinchuk, D [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1994-12-31

After a review of the main characteristics of the Atucha I nuclear power plant and its pressure vessel, the embrittlement surveillance capsules and the irradiation conditions are described; Charpy impact tests and tensile tests were performed on the irradiated samples, and results are discussed and compared to theoretical calculations: transition temperature shifts, displacement per atom values. 6 refs., 16 figs., 7 tabs.

Method of diagnosing a nuclear power plant

International Nuclear Information System (INIS)

Tamaoki, Tetsuo; Kawano, Koji

1983-01-01

Purpose : To exactly diagnose a nuclear power plant even upon occurrence of an abnormal state. Method : A statistical data calculation device prepares multidimensional vectors on every statistical amount for the unit of flowrate, pressure, temperature and neutron flux data in a nuclear power plant. A comparison-and-reference device compares them with the statistical amount rendered into multi-dimensional vectors corresponding to the generation of abnormality on every time in a normal operation for each of the fluctuation amount stored in a comparison and reference-value-store device. The diagnosis device performs diagnosis while using both of the standard pattern previously prepared and stored in the comparison and standard-value-store device and the pattern learnt and stored in the comparison and reference-value-store device. (Seki, T.)

Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

International Nuclear Information System (INIS)

Perrin, J.S.

1978-01-01

A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

Nuclear power plant outages

International Nuclear Information System (INIS)

1998-01-01

The Finnish Radiation and Nuclear Safety Authority (STUK) controls nuclear power plant safety in Finland. In addition to controlling the design, construction and operation of nuclear power plants, STUK also controls refuelling and repair outages at the plants. According to section 9 of the Nuclear Energy Act (990/87), it shall be the licence-holder's obligation to ensure the safety of the use of nuclear energy. Requirements applicable to the licence-holder as regards the assurance of outage safety are presented in this guide. STUK's regulatory control activities pertaining to outages are also described

Practical applications of probabilistic structural reliability analyses to primary pressure systems of nuclear power plants

International Nuclear Information System (INIS)

Witt, F.J.

1980-01-01

Primary pressure systems of nuclear power plants are built to exacting codes and standards with provisions for inservice inspection and repair if necessary. Analyses and experiments have demonstrated by deterministic means that very large margins exist on safety impacting failures under normal operating and upset conditions. Probabilistic structural reliability analyses provide additional support that failures of significance are very, very remote. They may range in degree of sophistication from very simple calculations to very complex computer analyses involving highly developed mathematical techniques. The end result however should be consistent with the desired usage. In this paper a probabilistic structural reliability analysis is performed as a supplement to in-depth deterministic evaluations with the primary objective to demonstrate an acceptably low probability of failure for the conditions considered. (author)

Assessment of EPRI water chemistry guidelines for new nuclear power plants

International Nuclear Information System (INIS)

Reid Richard; Kim Karen; McCree, Anisa; Eaker, Richard; Sawochka, Steve; Giannelli, Joe

2012-09-01

Water chemistry control technologies for nuclear power plants have been significantly enhanced over the past few decades to improve material and equipment reliability and fuel performance, and to minimize radionuclide production and transport. Chemistry Guidelines have been developed by the Electric Power Research Institute (EPRI) for currently operating plants and have been intermittently revised over the past twenty-five years for the protection of systems and components and for radiation management. As new plants are being designed for improved safety and increased power production, it is important to ensure that the designs consider implementation of state-of-the-art, industry developed water chemistry controls. In parallel, the industry will need to consider and update water chemistry guidelines as well as plant startup and operational strategies based on the advanced plant designs. EPRI has performed assessments of water chemistry control guidance or assumptions provided in design and licensing documents for several advanced plant designs. These designs include: Westinghouse AP1000 Pressurized Water Reactor AREVA US-EPR Pressurized Water Reactor Mitsubishi Nuclear Energy Systems/Mitsubishi Heavy Industries Advanced Pressurized Water Reactor Korea Hydro and Nuclear Power APR1400 Pressurized Water Reactor Toshiba Advanced Boiling Water Reactor (ABWR) General Electric-Hitachi Economic Simplified Boiling Water Reactor (ESBWR) The intent of these assessments was to identify key design differences in each of the new plant designs relative to the current operating fleet and to identify differences in water chemistry specifications or design assumptions provided in design and licensing documents for the plants in comparison to current EPRI Water Chemistry Guidelines. This paper provides a summary of the key results of these assessments. The fundamental design and operation of the advanced plants is similar to the currently operating fleet. As such, the new plants are

Application of Advanced Technology to Improve Plant Performance in Nuclear Power Plants

International Nuclear Information System (INIS)

Hashemian, H.M.

2011-01-01

providing the nuclear industry with the means to meet regulatory requirements, comply with technical specification provisions, or resolve operational and maintenance issues. Although OLM provides substantial benefits to the safety and economy of nuclear power plants, it is not widely used in the nuclear industry at this time for a number of reasons; the most important of which is regulatory constraints. In particular, the regulators must allow OLM to replace the conventional techniques for maintenance of safety-related equipment to make it worthwhile for utilities to retrofit their plants with OLM technologies. To this end, the U.S. Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) in the year 2000 accepting the OLM concept for condition-based calibration of safety-related pressure transmitters in nuclear power plants. However, according to the SER, each plant must still apply to the NRC and receive approval for OLM implementation if it is to be used in lieu of traditional calibration of safety-related equipment. This is, of course, a hindrance for OLM and has slowed its widespread use in the nuclear industry. As such, in the fall of 2008, representatives of the U.S. nuclear industry initiated an effort to obtain generic NRC licensing for the use of OLM in nuclear power plants. If approved, generic licensing will allow nuclear power plants to implement OLM without having to apply for an individual license for each plant. There is no doubt that this will incentivize the industry to proceed with OLM implementation at an accelerated rate. (author)

Coupling of RO-MSF hybrid desalination plants with nuclear reactors

International Nuclear Information System (INIS)

Al-Sulaiman, Khalil; Al-Mutaz, Ibrahim S.

1999-01-01

Full text.Reverse osmosis (RO) and multistage flash (MSF) desalination are the most widely commercial available processes. MSF utilizes stream in the brine heater as a primary source of energy. RO is derived mainly by electricity that pumps the feed water against the mambranes. Steam and electricity and be produced easily by nuclear reactors. Nuclear reactors may be coupled with deslination plants (MSF, RO or combined (hybrid) RO/MSF configuration). This integrated plant will be capable of producing power and water at reasonable cost. The capital and operating cost will be reduced and the excess power can be efficiently utilized. Maintenance and operating cost will drop significantly. In this paper, a techno-economic study of hybrid reverses osmosis /multistage flash desalination will be carried. The proposed configuration (hybrid RO/MSF) coupled with nuclear reactor is considered the most appropriate candidate system for the application of dual-purpose nuclear desalination plants. the design parameters for such a desalination hybrid system will be the applied pressure and recovery for reverse osmosis plant and the number of stages and the heat transfer areas for multistage flash plant

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

International Nuclear Information System (INIS)

2005-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

Contributions to economical and safe operation of nuclear power plants

International Nuclear Information System (INIS)

Ackermann, G.; Meyer, K.

1989-01-01

Selected results of scientific and technical research works in the Department 'Nuclear Power' of the Zittau Technical University are summarized which have been obtained on behalf of the Kombinat Kernkraftwerke 'Bruno Leuschner' and in conjunction with the education of scientific successors and have been partly adopted in textbooks. Works on improved utilization of nuclear fuel in pressurized water reactors are mentioned which, among other things, are related with the use of stretch-out mode of operation and optimization of nuclear fuel loading sequence. Results of experimental and theoretical investigations on coolant mixing in the reactor core are presented. A complex modelling of the dynamical long-term behaviour of nuclear power plants with pressurized water reactors due to xenon poisoning are briefly described. Finally, some results on noise diagnostics theory of power reactors are summarized. (author)

Nuclear power plants and house-building

International Nuclear Information System (INIS)

1982-04-01

In this report, it is studied where and under what conditions houses can be built in the neighbourhood of nuclear power plants. Building norms and location distances are investigated. First, the Dutch and foreign norms with respect to population densities are listed. Only industrial, densely populated neighbour countries are considered. Next, it is calculated what consequences for housing a nuclear station may have within a radius of 20 km. Using these calculations it is studied whether the existing Dutch requirements are satisfied. Because it is expected that the norms are likely to be tightened up under the pressure of the nuclear controversy, the existing Dutch situation is also compared with the tighter foreign norms. Finally, the results of the study are summarized and some conclusions are drawn. (Auth.)

Operational failure at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Szatmary, Z.

2003-01-01

NPP failures are ranked according to the International Nuclear Event Scale. To rank the failure first a presentation of the pressurized water plant is given, including fuel change, maintenance cleaning and decontamination process. The failure has been produced with fuel bars in the cleaning container. Consequences of the failure are small, negligible environmental pollution with radioactive material and significant financial outfall due to inactivity of block 2. Among the causes of the failure are design errors of the cleaning container, the pure chemical approach to cleaning, unknown risk factors for some of the cleaning staff, cleaning container has not been verified and approved by responsible authorities, the prevalence of economic and quantitative indicators of the plant on the detriment of safety. Organisational factors also contribute to the possibility of nuclear failures. Specialist training in Germany (where the container has been produced) is significantly reduced, while in Hungary the political tide has caused a permanent change in the higher echelons of the plant management, where nuclear specialists were not included. (Gy.M.)

Materialistic Aspects of Raising Resource of Pressurized Water Reactors for Low-Power Nuclear Plants

International Nuclear Information System (INIS)

Parshin, A.M.; Muratov, O.E.

2005-01-01

The opportunity of using ships reactors for low-power nuclear plants is considered. Some aspects of working constructional materials on cases of water-water reactors of ships nuclear units are considered. Advantages of raising resource of ships reactors are shown

Operation of Finnish nuclear power plants. Quarterly report, 3rd quarter 1996

International Nuclear Information System (INIS)

Sillanpaeae, T.

1997-02-01

Quarterly Reports on the operation of Finnish nuclear power plants describe events and observations relating to nuclear and radiation safety which the Finnish Centre for Radiation and Nuclear Safety (STUK) considers safety significant. Safety improvements at the plants are also described. The Report also includes a summary of the radiation safety of plant personnel and of the environment and tabulated data on the plants' production and load factors. In the third quarter of 1996, the Finnish nuclear power plant units were in power operation except for the annual maintenance outages of Loviisa plant units and a shutdown at Olkiluoto 1 to identify and repair malfunctions of a high pressure turbine control valve. The load factor average of all plant units was 77.2%. Events in the third quarter of 1996 were classified level 0 on the International Nuclear Event Scale. Occupational doses and radioactive releases off-site were below authorised limits. Radioactive substances were measurable in samples collected around the plants in such quantities only as have no bearing on the radiation exposure of the population. The names of Teollisuuden Voima Oy's plant units have changed. Olkiluoto 1 and Olkiluoto 2 now replace the names TVO I and TVO II previously used in quarterly reports. (orig.)

Analysis on capability of load following for nuclear power plants abroad and its enlightenment

Science.gov (United States)

Zheng, Kuan; Zhang, Fu-qiang; Deng, Ting-ting; Zhang, Jin-fang; Hao, Weihua

2017-01-01

With the acceleration adjustment of China’s energy structure, the development of nuclear power plants in China has been going back to the fast track. While as the trend of slowing electric power demand is now unmistakable, it enforces the power system to face much greater pressure in some coastal zones where the nuclear power plants are of a comparative big proportion, such as Fujian province and Liaoning province. In this paper, the capability of load following of nuclear power plants of some developed countries with high proportion of nuclear power generation such as France, US and Japan are analysed, also from the aspects including the safety, the economy and their practical operation experience is studied. The feasibility of nuclear power plants to participate in the peak regulation of system is also studied and summarized. The results of this paper could be of good reference value for the China’s nuclear power plants to participate in system load following, and also of great significance for the development of the nuclear power plants in China.

Assessment of possibility of primary water stress corrosion cracking occurrence based on residual stress analysis in pressurizer safety nozzle of nuclear power plant

International Nuclear Information System (INIS)

Lee, Kyoung Soo; Kim, W.; Lee, Jeong Geun

2012-01-01

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is required to generate PWSCC or what causes such high tensile stress. This study was performed to predict the magnitude of weld residual stress and operating stress and compare it with previous experimental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by numerical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up analysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mockup. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

Prestressed reactor vessel for nuclear power plants

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1982-01-01

With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

Assessment and Management of ageing of major nuclear power plant components important to safety: PWR pressure vessels

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g., caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs; and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which involves the integration of

Nuclear power plant siting

International Nuclear Information System (INIS)

Sulkiewicz, M.; Navratil, J.

The construction of a nuclear power plant is conditioned on territorial requirements and is accompanied by the disturbance of the environment, land occupation, population migration, the emission of radioactive wastes, thermal pollution, etc. On the other hand, a nuclear power plant makes possible the introduction of district heating and increases the economic and civilization activity of the population. Due to the construction of a nuclear power plant the set limits of negative impacts must not be exceeded. The locality should be selected such as to reduce the unfavourable effects of the plant and to fully use its benefits. The decision on the siting of the nuclear power plant is preceded by the processing of a number of surveys and a wide range of documentation to which the given criteria are strictly applied. (B.H.)

Economic feasibility of heat supply from simple and safe nuclear plants

International Nuclear Information System (INIS)

Tian, J.

2001-01-01

Use of nuclear energy as a heating source is greatly challenged by the economic factor since the nuclear heating reactors have relative small size and often the lower plant load factor. However, use of very simple reactor could be a possible way to economically supply heat. A deep pool reactor (DPR) has been designed for this purpose. The DPR is a novel design of pool type reactor for heat only supply. The reactor core is put in a deep pool. By only putting light static water pressure on the core coolant, the DPR will be able to meet the temperature requirements of heat supply for district heating. The feature of simplicity and safety of DPR makes a decrease of investment cost compared to other reactors for heating only purposes. According to the economical assessments, the capital investment to build a DPR plant is much less than that of a pressurized reactor with pressure vessels. For the DPR with 120 or 200 MW output, it can bear the economical comparison with a usual coal-fired heating plant. Some special means taken in DPR design make an increase of the burn-up level of spent fuel and a decrease of fuel cost. The feasibility studies of DPR in some cities in China show that heating cost using nuclear energy is only one third of that by coal and only one tenth of that by nature gas. Therefore, the DPR nuclear heating system provides an economically attractive solution to satisfy the demands of district heating without contributing to increasing greenhouse gas emissions

Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

International Nuclear Information System (INIS)

Wati, Nurokhim

2008-01-01

Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

Images of nuclear power plants

International Nuclear Information System (INIS)

Hashiguchi, Katsuhisa; Misumi, Jyuji; Yamada, Akira; Sakurai, Yukihiro; Seki, Fumiyasu; Shinohara, Hirofumi; Misumi, Emiko; Kinjou, Akira; Kubo, Tomonori.

1995-01-01

This study was conducted to check and see, using Hayashi's quantification method III, whether or not the respondents differed in their images of a nuclear power plant, depending on their demographic variables particularly occupations. In our simple tabulation, we compared subject groups of nuclear power plant employees with general citizens, nurses and students in terms of their images of a nuclear power plant. The results were that while the nuclear power plant employees were high in their evaluations of facts about a nuclear power plant and in their positive images of a nuclear power plant, general citizens, nurses and students were overwhelmingly high in their negative images of a nuclear power plant. In our analysis on category score by means of the quantification method III, the first correlation axis was the dimension of 'safety'-'danger' and the second correlation axis was the dimension of 'subjectivity'-'objectivity', and that the first quadrant was the area of 'safety-subjectivity', the second quadrant was the area of 'danger-subjectivity', the third quadrant as the area of 'danger-objectivity', and the forth quadrant was the area of 'safety-objectivity'. In our analysis of sample score, 16 occupation groups was compared. As a result, it was found that the 16 occupation groups' images of a nuclear power plant were, in the order of favorableness, (1) section chiefs in charge, maintenance subsection chiefs, maintenance foremen, (2) field leaders from subcontractors, (3) maintenance section members, operation section members, (4) employees of those subcontractors, (5) general citizens, nurses and students. On the 'safety-danger' dimension, nuclear power plant workers on the one hand and general citizens, nurses and students on the other were clearly divided in terms of their images of a nuclear power plant. Nuclear power plant workers were concentrated in the area of 'safety' and general citizens, nurses and students in the area of 'danger'. (J.P.N.)

Criteria for safety-related nuclear-power-plant operator actions: 1982 pressurized-water-reactor (PWR) simulator exercises

International Nuclear Information System (INIS)

Crowe, D.S.; Beare, A.N.; Kozinsky, E.J.; Haas, P.M.

1983-06-01

The primary objective of the Safety-Related Operator Action (SROA) Program at Oak Ridge National Laboratory is to provide a data base to support development of criteria for safety-related actions by nuclear power plant operators. When compared to field data collected on similar events, a base of operator performance data developed from the simulator experiments can then be used to establish safety-related operator action design evaluation criteria, evaluate the effects of performance shaping factors, and support safety/risk assessment analyses. This report presents data obtained from refresher training exercises conducted in a pressurized water reactor (PWR) power plant control room simulator. The 14 exercises were performed by 24 teams of licensed operators from one utility, and operator performance was recorded by an automatic Performance Measurement System. Data tapes were analyzed to extract operator response times (RTs) and error rate information. Demographic and subjective data were collected by means of brief questionnaires and analyzed in an attempt to evaluate the effects of selected performance shaping factors on operator performance

Noise diagnosis - a method for early detection of failures in a nuclear plant

International Nuclear Information System (INIS)

Brinckmann, H.F.

1981-01-01

Noise diagnosis constitutes one method for early detection of plant failures. The method is based on the fact that nearly all undesired processes in a nuclear power plant make a measurable contribution to the noise portion of signals. Well-known examples of undesired processes in pressurized water reactors include core-barrel movement, the vibration of control elements, the appearance of loose parts in the coolant flow, and the process of coolant boiling. Each of these processes has been implicated in past nuclear plant failures. In the German Democratic Republic (GDR) P. Liewers and his colleagues have introduced noise analysis systems into the primary circuit of WWER-440 pressurized water reactors (PWR). The most progressive version (RAS-II) has become a prototype for research and routine investigations. This system is described. (author)

NuPEER Dijon 2005 Symposium. Ageing issues in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Emond, David (ed.) [BCCN, Autorite de Surete Nucleaire, ASN, 6, place du Colonel Bourgoin, 75572 Paris Cedex 12 (France)

2005-07-01

The French Nuclear Safety Authority (ASN) organized an international symposium on regulatory aspects of ageing issues for nuclear pressure equipment. The ageing of nuclear pressure equipment is an issue of growing importance for nuclear regulators and material experts worldwide as age-related degradation of major pressure-retaining components challenges the remaining operating life of nuclear power plants. This symposium aimed at providing a forum for technical exchange among the staffs responsible for nuclear pressure equipment within the safety authorities and the associated expertise organisations. The contents of the symposium is as follows: 1. Control and supervision of safety of nuclear pressure equipment in France and abroad; 1.1. Position of the French Nuclear Safety Authority (1 paper); 1.2. Regulatory practices worldwide (4 papers); 1.3. Licence renewal: Field experience (2 papers); 1.4. Role of international organisations (1 paper); 2. Management of equipment and materials: From design to degradation mechanisms; 2.1. Operation and equipment (4 papers); 2. Evolution of materials (4 papers); 2.3. Fatigue degradation mechanisms (3 papers); 2.4. Contribution of research and development (4 papers); 3. In-service inspection: Evolutions, methods and strategies; 3.1. Methods and evolution (1 paper); 3.2. Qualification of methods (2 papers); 3.3. Surveillance strategies (2 papers); 4. Testimonies and points of view of utilities (3 papers); 5. Ageing issues taken into account in non nuclear fields (2 papers). The symposium began with workshops devoted to: Operation and equipment; Behaviour of materials; Fatigue degradations; Contributions of research and development. The symposium continued with plenary session that addressed the following issues: Control and supervision of safety of nuclear pressure equipment; Role of international organisations; In-service inspection: Objectives, methods and strategies; Point of view of utilities; Technical summary and

76 FR 73720 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000...

Science.gov (United States)

2011-11-29

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0272] Knowledge and Abilities Catalog for Nuclear Power...) is issuing for public comment a draft NUREG, NUREG-2103, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000 Pressurized-Water Reactors. DATES: Submit...

Introduction to the 'CAS' nuclear propulsion plant for ships: specific safety options

International Nuclear Information System (INIS)

Verdeau, J.J.; Baujat, J.

1978-01-01

After a brief review of the development of nuclear propulsion in FRANCE (Land Based Prototype PAT 1964 - Navy nuclear ships - Advanced Nuclear Boiler Prototype CAP 1975 and now the CAS nuclear plant), the specific safety options of CAS are presented: cold, compartmented fuel (plates); reduced flow during LOCA; permanent cooling of fuel during LOCA; pressurized, entirely passive containment; no control rod ejection and possibility of temporary storage of spent fuel on board [fr

Impact of power uprate on environmental qualification of equipment in nuclear power plants

International Nuclear Information System (INIS)

Raheja, R.D.; Mohiuddin, A.; Alsammarae, A.

1996-01-01

Many nuclear power facilities are finding it economically beneficial to increase reactor output, from operating plants, by resorting to power uprates. A power uprate implies that a utility can increase the reactor output, or the megawatts generated, by increasing steam pressure without adding or changing any plant systems. This is perhaps one of the least expensive options for increasing the generating capacity of a power plant. However, a nuclear plant requires a comprehensive review of the plant systems, structures and components to assure their capability to withstand the resulting increased normal and accident plant conditions. A power uprate will typically result in a plant operating at higher than the originally designed environmental conditions. Safety related equipment in nuclear plants is presently qualified to the UFSAR Chapter 15 accident events and the resulting temperatures, pressures, radiation levels etc. These values will increase when the reactor is producing a higher MWe output. Components that are sensitive to the environment must be re-evaluated and assessed to determine their acceptability and operability under the revised environmental conditions. Most safety-related mechanical and electrical equipment will require an assessment from an environmental qualification standpoint. Utilities must perform this task in a systematic, auditable and cost effective manner to optimize their resources and minimize plant costs associated with modifications, replacements or equipment testing. This paper discusses various approaches and provides recommendations to achieve equipment qualification while satisfying the plant's objective of a power uprate

Heat recovery from nuclear power plants

International Nuclear Information System (INIS)

Safa, H.

2012-01-01

The thermodynamic efficiency of a standard Nuclear Power Plant (NPP) is around 33%. Therefore, about two third of the heat generated by the nuclear fuel is literally wasted in the environment. Given the fact that the steam coming out from the high pressure turbine is superheated, it could be advantageously used for non electrical applications, particularly for district heating. Considering the technological improvements achieved these last years in heat piping insulation, it is now perfectly feasible to envisage heat transport over quite long distances, exceeding 200 km, with affordable losses. Therefore, it could be energetically wise to revise the modifications required on present reactors to perform heat extraction without impeding the NPP operation. In this paper, the case of a French reactor is studied showing that a large fraction of the wasted nuclear heat can be actually recovered and transported to be injected in the heat distribution network of a large city. Some technical and economical aspects of nuclear district heating application are also discussed. (author)

Health protection and industrial safety. Nuclear power plants

International Nuclear Information System (INIS)

1987-03-01

The standard applies to primary circuit components including its auxiliary facilities, and of the secondary circuit of nuclear power plants with pressurized water reactors; to lifting gear and load take-ups for the transport of nuclear fuel and primary circuit components; to elevators within the containment, and to electrical installations. Part 3 specifies the behaviour of workers in conformity with safety provisions during operation, inspection, lifetime surveillance, functional testing, and maintenance. Special demands are made on the water regime and on elevators, lifting gear, and load take-ups

Application of analysis technology in nuclear plant

International Nuclear Information System (INIS)

Takaoka, Keiko; Miura, Hiromi; Umeda, Kenji

1996-01-01

Recently, thanks to the rapid improvement of EWS performance, the authors have been able to carry out design evaluation comparatively, easily, utilizing computational fluid dynamics (CFD) technology. The Nuclear Plant Engineering Department has carried out some analyses in the past several years with the main purpose of evaluating the design of nuclear reactor internals. These studies included ''Thermal Hydraulic Analysis for Top Plenum'' and ''Flow Analysis for Lower Plenum''. It is considered to be a special matter in thermal hydraulic analysis of the top plenum that temperature distribution has been estimated with a relatively small number of meshes by means of an imaginary spray nozzle, and in the flow analysis for the lower plenum that flow distribution has been found to change largely, depending on the reactor internals. One of the ways to confirm the safety of nuclear plants, detailed structural analysis, is required for all possible combinations of transient and load conditions during operation. In particular, it is very important to clarify the thermal stress behavior under operating conditions and to evaluate fatigue analysis in accordance with the Code Requirements. However, it is very complicated and it takes a lot of time. A new system was developed which can operate continuously all of the definitions of the analytical model, the analyzation of pressurized thermal and external stress, and editing reports. In this paper, the authors introduce this system and apply it to a pressurized water reactor

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

2003-01-01

This 2003 version of Elecnuc contents information, data and charts on the nuclear power plants in the world and general information on the national perspectives concerning the electric power industry. The following topics are presented: 2002 highlights; characteristics of main reactor types and on order; map of the French nuclear power plants; the worldwide status of nuclear power plants on 2002/12/3; units distributed by countries; nuclear power plants connected to the Grid by reactor type groups; nuclear power plants under construction; capacity of the nuclear power plants on the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear plants by country at the end 2002; performance indicator of french PWR units; trends of the generation indicator worldwide from 1960 to 2002; 2002 cumulative Load Factor by owners; nuclear power plants connected to the grid by countries; status of license renewal applications in Usa; nuclear power plants under construction; Shutdown nuclear power plants; exported nuclear power plants by type; exported nuclear power plants by countries; nuclear power plants under construction or order; steam generator replacements; recycling of Plutonium in LWR; projects of MOX fuel use in reactors; electricity needs of Germany, Belgium, Spain, Finland, United Kingdom; electricity indicators of the five countries. (A.L.B.)

Blow-off device for limiting excess pressure in nuclear power plants, especially in boiling-water nuclear power plants

International Nuclear Information System (INIS)

Kuehnel, R.

1979-01-01

In a blow-off device for limiting excess pressure in nuclear power plants, at least one condensation tube disposed so that a lower outlet end thereof is immersed in a volume of water in a condensation chamber having a gas cushion located in a space above the volume of water, and the upper inlet end of the condensation tube extending out of the volume of water and being connectible to a source of steam that is to be condensed or a steam-air mixture, the outlet end of the condensation tube, for smoothing the condensation, being provided with wall parts forming passages extending in axial direction, delimited from one another and terminating in the water volume, the wall parts serving to subdivide steam flow from the source thereof and bubbles produced thereby in the water volume, the wall parts being constructed as a tube attachment and being formed with an opening corresponding to the outlet end of the condensation tube and by means of which the tube attachment is mounted on the outlet end of the condensation tube, a first group of the wall parts in the tube attachment being disposed in alignment with the outlet end of the condensation tube, and a second group of the wall parts surrounding the first group thereof, the passages formed by the second group of the wall parts communicating laterally with the passages formed by the first group of the wall parts, the passages formed by the second group of the wall parts, at least at the upper ends thereof, communicating with the water volume

Operating reliability of valves in French pressurized water nuclear power plants

International Nuclear Information System (INIS)

Conte

1986-10-01

Taking into account the large numbers of valves (about 10000) of a PWR nuclear power plant, the importance of some valves in the safety functions and the cost resulting from their unavailability, the individual operability of these equipments has to be ensured at a high reliability level. This assurance can be obtained by means of an effort at all the stages which contribute to the quality of the product: design, qualification tests, fabrication, tests at the start-up stage, maintenance and tests during the power plant operation, experience feedback. This paper emphasizes more particularly on the tests carried out on loops of qualification [fr

Technology and costs for decommissioning of Swedish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-06-01

The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs.

Technology and costs for decommissioning of Swedish nuclear power plants

International Nuclear Information System (INIS)

1994-06-01

The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs

Development of a dynamic model of a BWR nuclear power plant

International Nuclear Information System (INIS)

Nonboel, E.

1975-12-01

A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

Actions concerning nuclear power plant life evaluation

International Nuclear Information System (INIS)

Chocron, M.; Fabbri, S.; Mizrahi, R.; Savino, E.J.; Versaci, R.A.

1998-01-01

One of the main activities to be undertaken by CNEA will be to provide technological assistance to NASA in problems concerning NPP operation. Works on life extensions of NPP are included in these activities. To fulfill these requirements the Atomic Energy National Commission (CNEA) has constituted a technical committee for Nuclear Power Plants Support (CAPCEN). CAPCEN should be the knowledge reservoir of those issues concerning the performance, safety and life extension of Nuclear Power Plants. One of CAPCEN's most important activities is to promote research work connected with such issues. The main technical areas are: Pressure Vessel and Piping, Heat Exchanges and Fuel Channels and Reactor Inner Components. Efforts are focused on the identification of the main components susceptible of ageing, the study of their ageing mechanisms, the follow-up of their behaviour during operation, and the measures taken to extend their life. (author)

76 FR 1469 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

Science.gov (United States)

2011-01-10

... Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Environmental Assessment... Plant, LLC, the licensee, for operation of the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2... Impact Statement for License Renewal of Nuclear Plants, Calvert Cliffs Nuclear Power Plant (NUREG-1437...

Operation of Finnish nuclear power plants. Quarterly report, 3rd quarter 1997

International Nuclear Information System (INIS)

Tossavainen, K.

1998-04-01

Quarterly reports on the operation of Finnish nuclear power plants describe events and observations relating to nuclear and radiation safety that the Radiation and Nuclear Safety Authority of Finland (STUK) considers safety significant. Safety improvements at the plants are also described. The Report also includes a summary of the radiation safety of plant personnel and of the environment and tabulated data on the plants' production and load factors. The Finnish nuclear power plant units were in power operation in the third quarter of 1997, except for the annual maintenance outages of Loviisa plant units which lasted well over a month in all. There was also a brief interruption in electricity generation at Olkiluoto 1 for repairs and at Olkiluoto 2 due to a disturbance at the turbine plant. All plant units were in long-term test operation at upgraded reactor power level approved by STUK. The load factor average of all plant units was 87.6 %. One event in the third quarter was classified level 1 on the International Nuclear Event Scale (INES). It was noted at Loviisa 2 that one of four pressurized water tanks in the plant unit's emergency cooling system had been inoperable for a year. Other events in this quarter were INES level 0. Occupational doses and radioactive releases off-site were below authorized limits. Radioactive substances were measurable in samples collected around the plants in such quantities only as have no bearing on the radiation exposure of the population. (orig.)

Systematic generation of rules for nuclear power plant diagnostics

International Nuclear Information System (INIS)

Reifman, J.; Lee, J.C.

1988-01-01

The knowledge base of an expert system is generally represented by a set of heuristic rules derived from the expert's own experience and judgmental knowledge. These heuristic or production rules are cast as if (condition), then (consequence) statements, and represent, for nuclear power plant diagnostic systems, information connecting symptoms to failures. In this paper, the authors apply an entropy minimax pattern recognition algorithm to automate the process of extracting and encoding knowledge into a set of rules. Knowledge is extracted by recognizing patterns in plant parameters or symptoms associated with failures or transient events, and is encoded by casting the discovered patterns as production rules. The paper discusses how the proposed method can systematically generate rules that characterize failure of pressurizer components based on transient events analyzed with a pressurizer components based on transient events analyzed with a pressurizer water reactor simulator program

Internal exposure monitoring of personnel of a nuclear power plant with pressurized-water reactors

International Nuclear Information System (INIS)

Krueger, F.W.; Poulheim, K.F.; Rueger, G.; Schreiter, W.D.

1982-01-01

In the GDR a programme for monitoring the internal radiation exposure of personnel has been established in the Bruno Leuschner Nuclear Power Plant, Greifswald, which allows one to estimate the effective dose equivalent in the way recommended by the ICRP. The measuring equipment used, and the methods of calibration and of evaluation of results are described. At present about 400 persons are monthly monitored with a thorax monitor in the nuclear power plant. If an investigation level - corresponding to an effective dose equivalent of 0.3mSv/month - is exceeded, a more exact measurement is made in the whole-body counter at the National Board for Nuclear Safety and Radiation Protection of the GDR. In addition, a selected group of 50 persons is measured twice yearly in the whole-body counter. The measurements show the high effectiveness of the protective measures against radionuclide intake by workers in the nuclear power plant, resulting in a contribution of less than 1% to the collective dose of the personnel. A correlation has been found between external and internal exposure indicating that, in general, there will be a higher intake only under conditions resulting also in higher external exposures. The highest individual values of internal exposure found are below 0.5mSv/month and thus within the range of the lower detection limit of dosimeter films used for monitoring the external exposure. (author)

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World's First nuclear power plant

Science.gov (United States)

Rachkov, V. I.; Kalyakin, S. G.; Kukharchuk, O. F.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

Successful commissioning in the 1954 of the World's First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact

Technical feasibility and costs of the retention of radionuclides during accidents in nuclear power plants demonstrated by the example of a pressurized water reactor

International Nuclear Information System (INIS)

Braun, H.; Grigull, R.; Lahner, K.; Gutowski, H.; Weber, J.

1985-01-01

The maximum allowable radiation doses during accidents in nuclear power plants, i.e., 5 rem whole-body dose and 15 rem thyroid dose, have been laid down in the German Radiation Protection Act. In order to ensure that these limits are not exceeded for all exposure paths including the ingestion path or, if possible, to remain far below them, the Federal Ministry of the Interior has initiated a study on the effectiveness and cost of additional safety features for reducing the release of activity and the dose exposure during accidents in nuclear power plants. Detailed investigations were carried out for the following three radiologically representative types of accidents: break of a reactor coolant line, break of an instrument line in one of the outer ring rooms, and break of a main stream line outside the containment. The technical basis of the study was a BBR-type nuclear power plant with pressurized water reactor and once-through steam generator. I-131 was chosen for determining the activity release as this is the critical nuclide for the ingestion path. Altogether 33 feasible technical measures were investigated and their potential improvement was assessed

Nuclear power plant with a safety enclosure

International Nuclear Information System (INIS)

Keller, W.; Krueger, J.; Ropers, J.; Schabert, H.P.

1976-01-01

A nuclear power plant has a safety enclosure for a nuclear reactor. A fuel element storage basin is also located in this safety enclosure and a fuel element lock extends through the enclosure, with a cross-sectional size proportioned for the endwise passage of fuel elements, the lock including internal and external valves so that a fuel element may be locked endwise safely through the lock. The lock, including its valves, being of small size, does not materially affect the pressure resistance of the safety enclosure, and it is more easily operated than a lock large enough to pass people and fuel element transport vessels

Enhancing leadership at a nuclear power plant - a systematic approach

International Nuclear Information System (INIS)

Jupiter, P.

1989-01-01

The increasing use of advanced technology, greater regulatory oversight, and critical public scrutiny create numerous pressures for leaders within nuclear power plant systems (NPPSs). These large, complex industrial installations have unusually high expectations imposed for safety and efficiency of operation - without the luxury of trial-and-error learning. Industry leaders assert that enhanced leadership and management can substantially improve the operating performance of a nuclear power plant. The need has been voiced within the nuclear industry for systematic and effective methods to address leadership and management issues. This paper presents a step-by-step model for enhancing leadership achievement within NPPS, which is defined as the combined structural, equipment, and human elements involved in a plant's operation. Within the model, key areas for which the leader is responsible build upon each other in sequential order to form a solid strategic structure; teachable actions and skills form an ongoing cycle for leadership achievement. Through the model's continued and appropriate functioning, a NPPS is likely to maintain its viability, productivity, and effectiveness for the full licensed term of a plant

Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

International Nuclear Information System (INIS)

Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

2015-01-01

In ICONE22, the authors presented the Integrated Head Assembly (IHA) design concept implemented at Callaway Nuclear Power Plant in Missouri, USA. The IHA concept is implemented to reduce the outage duration and the associated radiation exposure to the workers by reducing critical path time during Plant Refueling Outage. One of the head area components in the IHA is a steel missile shield designed to protect the Control Rod Drive Mechanism (CRDM) assembly from damaging other safety-related components in the vicinity in the Containment. Per Federally implemented General Design Criteria for commercial nuclear plants in the USA, the design of Nuclear Steam Supply System (NSSS) must provide protection from the damages caused by a postulated event of CRDM housing units and their associated parts disengaging from the reactor vessel assembly. This event is considered as a Loss of Coolant Accident (LOCA) and assumes that once the CRDM housing unit and their associated parts disengage from the reactor vessel internals assembly, they travel upward by the water jet with the following sequence of events: Per Reference 1, the drive shaft and control rod cluster are forced out of the reactor core by the differential pressure across the drive shaft with the assumption that the drive shaft and control rod cluster, latched together, are fully inserted when the accident occurs. After the travel, the rod cluster control spider will impact the lower side of the upper support plate inside the reactor vessel fracturing the flexure arms in the joint freeing the drive shaft from the control rod cluster. The control rod cluster is stopped by the upper support plate and will remain below the upper support plate during this accident. However, the drive shaft will continue to accelerate in the upward direction until it is stopped by a safety feature in the IHA. The integral missile shield as a safety feature in the IHA is designed to stop the CRDM drive shaft from moving further up in the

MARS, 600 MWth NUCLEAR POWER PLANT

International Nuclear Information System (INIS)

Cumo, M.; Naviglio, A.; Sorabella, L.

2004-01-01

MARS (Multipurpose Advanced Reactor, inherently Safe) is a 600 MWth, single loop, pressurized light water reactor (PWR), developed at the Dept. of Nuclear Engineering and Energy Conversion of the University of Rome ''La Sapienza''. The design was focused to a multipurpose reactor to be used in high population density areas also for industrial heat production and, in particular, for water desalting. Using the well-proven technology and the operation experience of PWRs, the project introduces a lot of innovative features hugely improving the safety performance while keeping the cost of KWh competitive with traditional large power plants. Extensive use of passive safety, in depth plant simplification and decommissioning oriented design were the guidelines along the design development. The latest development in the plant design, in the decommissioning aspects and in the experimental activities supporting the project are shown in this paper

Life management plants at nuclear power plants PWR

International Nuclear Information System (INIS)

Esteban, G.

2014-01-01

Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

Benchmarking Nuclear Power Plants

International Nuclear Information System (INIS)

Jakic, I.

2016-01-01

One of the main tasks an owner have is to keep its business competitive on the market while delivering its product. Being owner of nuclear power plant bear the same (or even more complex and stern) responsibility due to safety risks and costs. In the past, nuclear power plant managements could (partly) ignore profit or it was simply expected and to some degree assured through the various regulatory processes governing electricity rate design. It is obvious now that, with the deregulation, utility privatization and competitive electricity market, key measure of success used at nuclear power plants must include traditional metrics of successful business (return on investment, earnings and revenue generation) as well as those of plant performance, safety and reliability. In order to analyze business performance of (specific) nuclear power plant, benchmarking, as one of the well-established concept and usual method was used. Domain was conservatively designed, with well-adjusted framework, but results have still limited application due to many differences, gaps and uncertainties. (author).

The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

International Nuclear Information System (INIS)

Huber, Horacio

1989-01-01

Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

Integrated approach to optimize operation and maintenance costs for operating nuclear power plants

International Nuclear Information System (INIS)

2006-06-01

In the context of increasingly open electricity markets and the 'unbundling' of generating companies from former utility monopolies, an area of major concern is the economic performance of the existing fleet of nuclear power plants. Nuclear power, inevitably, must compete directly with other electricity generation sources. Coping with this competitive pressure is a challenge that the nuclear industry should meet if the nuclear option is to remain a viable one. This competitive environment has significant implications for nuclear plant operations, including, among others, the need for the more cost effective management of plant activities, and the greater use of analytical tools to balance the costs and benefits of proposed activities, in order to optimize operation and maintenance costs, and thus insure the economic competitiveness of existing nuclear power plants. In the framework of the activities on Nuclear Economic Performance Information System (NEPIS), the IAEA embarked in developing guidance on optimization of operation and maintenance costs for nuclear power plants. The report was prepared building on the fundamental that optimization of operation and maintenance costs of a nuclear power plant is a key component of a broader integrated business strategic planning process, having as overall result achievement of organization's business objectives. It provides advice on optimization of O and M costs in the framework of strategic business planning, with additional details on operational planning and controlling. This TECDOC was elaborated in 2004-2005 in the framework of the IAEA's programme on Nuclear Power Plant Operating Performance and Life Cycle Management, with the support of two consultants meetings and one technical meeting and based on contributions provided by participants. It can serve as a useful reference for the management and operation staff within utilities, nuclear power plant operators and regulators and other organizations involved in

Developing an economic performance system to enhance nuclear power plant competitiveness

International Nuclear Information System (INIS)

2002-01-01

In 1999 about 16% of the world's electricity was produced by nuclear power, and the total worldwide operating experience of nuclear power plants was over 9200 reactor-years. Some 16 countries are dependent on nuclear power for more than 25% of their electricity generation. In some countries deregulation of the electricity market has either happened or is currently ongoing, while in others it is planned for the future. Nevertheless, many countries are already facing open electricity markets and operating costs are under unprecedented pressure, with competition expected to come soon to the nuclear industry worldwide. To a certain extent, however, the industry has already prepared or is currently preparing to face competition. This report is primarily intended for nuclear power plant and utility managers. It discusses the means and principal issues for the development of the nuclear economic performance international system (NEPIS), which should enhance nuclear power plant competitiveness. The following issues are addressed: The major transformations occurring in the electricity generation industry that require reductions in operations and maintenance costs at nuclear utilities; The methods that nuclear plant management use to identify and justify the economic optimum level of a plant and its use of resources; The value of collecting cost and performance data and the analysis techniques that use that data; The cost data required to be collected; The difficulty of collecting data with existing cost accounting systems; The new cost accounting and collection systems that will be required, The cost effectiveness of the overall process. This report also presents the preliminary results of a pilot project that was established to collect cost data on a few nuclear power plants and was used to verify the adequacy of the definitions and terminology set for NEPIS

Should the service life of nuclear power plants be limited for technical reasons?

International Nuclear Information System (INIS)

Schomer, E.

1995-01-01

German nuclear power plant operating licenses take into account the fact that the design lifetime of the facilities is 40 years. The technological lifetime, however, is governed by actual ageing processes and operational burden. The service life of nuclear power plants can be extended by replacement of the steam generator and by repair or replacement of the reactor pressure vessel or parts thereof. Attention is also paid to the reactor protection systems. (J.B.). 8 figs

Analysis of failed nuclear plant components

International Nuclear Information System (INIS)

Diercks, D.R.

1993-01-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power-generating stations since 1974. The considerations involved in working with an analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (1) intergranular stress-corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor

Analysis of failed nuclear plant components

International Nuclear Information System (INIS)

Diercks, D.R.

1992-07-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power generating stations since 1974. The considerations involved in working with and analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (a) intergranular stress corrosion cracking of core spray injection piping in a boiling water reactor, (b) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressure water reactor, (c) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (d) failure of pump seal wear rings by nickel leaching in a boiling water reactor

Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels. 2007 update

International Nuclear Information System (INIS)

2007-06-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1120 documented ageing assessment and management practices for pressurized water reactor (PWR) reactor pressure vessels (RPVs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. primary water stress corrosion cracking (PWSCC) of Alloy 600 control rod drive mechanism (CRDM) penetrations and boric acid corrosion/wastage of RPV heads, which threatened the integrity of the RPV heads. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1120 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update IAEA-TECDOC-1120 in order to provide current ageing management guidance for PWR RPVs to all involved in the operation and regulation of PWRs and thus to help ensure PWR RPV integrity in IAEA Member States throughout their entire service life

Bruce Power's nuclear pressure boundary quality assurance program requirements, implementation and transition

International Nuclear Information System (INIS)

Krane, J.C.

2009-01-01

The development of a full scope nuclear pressure boundary quality assurance program in Canada requires extensive knowledge of the structure and detailed requirements of codes and standards published by the Canadian Standards Association (CSA) and American Society of Mechanical Engineers (ASME). Incorporation into company governance documents and implementation of these requirements while managing the transition to more recent revisions of these codes and standards represents a significant challenge for Bruce Power, Canada's largest independent nuclear operator. This paper explores the key developments and innovative changes that are used to ensure successful regulatory compliance and effective implementation of the Bruce Power Pressure Boundary Quality Assurance Program. Challenges and mitigating strategies to sustain this large compliance based program at Bruce Power's 8 unit nuclear power plant site will also be detailed. (author)

Impact of external grid disturbances on nuclear power plants; Rueckwirkungen von Netzstoerungen auf Kernkraftwerke

Energy Technology Data Exchange (ETDEWEB)

Arains, Robert; Arnold, Simone; Brueck, Benjamin; Mueller, Christian; Quester, Claudia; Sommer, Dagmar

2017-06-15

The electrical design of nuclear power plants and the reliability of their electrical power supply including the offsite power supply are of high importance for the safe operation of the plants. The operating experience of recent years has shown that disturbances in the external grid can have impact on the electrical equipment of nuclear power plants. In the course of this project, possible causes and types of grid disturbances were identified. Based on these, scenarios of grid disturbances were developed. In order to investigate the impact of the developed scenarios of grid disturbances on the electrical equipment of nuclear power plants, the auxiliary power supply of a German pressurized water reactor of type Konvoi was simulated using the simulation tool NEPLAN. On the basis of the results of the analyses, it was identified whether there are possible measures to prevent the spread of grid disturbances in the plants which have not been implemented in the nuclear power plants today.

Prospect and potential of nuclear power plants in Indonesia

International Nuclear Information System (INIS)

Subki, I.R.; Adiwardojo; Kasim, M.S.; Iskandar, A.; Mulyanto

1997-01-01

In line with the national energy policy of Indonesia in promoting the intensification, diversification and conversation of energy, some important steps need to be taken in order to establish alternative energy strategies which will be decisive in the formulation and development of the national energy plan in the future. At present, Indonesia does not have any nuclear power plants. The introduction of nuclear power In Indonesia is not only to reach an optimum energy mix based on costs and the environment, but also to relieve the pressure arising from increasing domestic demand for oil and gas. This paper addresses the present feasibility study being performed on the introduction of nuclear power plants in Indonesia. It is anticipated that nuclear power will contribute about 10% of Indonesia's electrical supply as of the year 2019. This represents approximately 12,600 MWe in capability. The paper describes the results, to date, of the Feasibility Study on nuclear power including the national energy market analysis, the electricity expansion plan and the associated role of nuclear power, the economics and financial plan, site studies on volcanology, seismology and the environment. (author). 3 refs, 1 fig., 4 tabs

Fuel handling and storage systems in nuclear power plants

International Nuclear Information System (INIS)

1984-01-01

The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

Piping engineering for nuclear power plant

International Nuclear Information System (INIS)

Curto, N.; Schmidt, H.; Muller, R.

1988-01-01

In order to develop piping engineering, an adequate dimensioning and correct selection of materials must be secured. A correct selection of materials together with calculations and stress analysis must be carried out with a view to minimizing or avoiding possible failures or damages in piping assembling, which could be caused by internal pressure, weight, temperature, oscillation, etc. The piping project for a nuclear power plant is divided into the following three phases. Phase I: Basic piping design. Phase II: Final piping design. Phase III: Detail engineering. (Author)

Are atomic power plants saver than nuclear power plants

International Nuclear Information System (INIS)

Roeglin, H.C.

1977-01-01

It is rather impossible to establish nuclear power plants against the resistance of the population. To prevail over this resistance, a clarification of the citizens-initiatives motives which led to it will be necessary. This is to say: It is quite impossible for our population to understand what really heappens in nuclear power plants. They cannot identify themselves with nuclear power plants and thus feel very uncomfortable. As the total population feels the same way it is prepared for solidarity with the citizens-initiatives even if they believe in the necessity of nuclear power plants. Only an information-policy making transparent the social-psychological reasons of the population for being against nuclear power plants could be able to prevail over the resistance. More information about the technical procedures is not sufficient at all. (orig.) [de

Distributed control system for CANDU 9 nuclear power plant

International Nuclear Information System (INIS)

Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

1996-01-01

Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. The CANDU 9 plant monitoring, annunciation, and control functions are implemented in two evolutionary systems; the distributed control system (DCS) and the plant display system (PDS). The CDS implements most of the plant control functions in a single hardware platform. The DCS communicates with the PDS to provide the main operator interface and annunciation capabilities of the previous control computer designs along with human interface enhancements required in a modern control system. (author)

Simulation study of a system for diagnosis of nuclear power plant operation

International Nuclear Information System (INIS)

Wakabayashi, J.; Fukumoto, A.

1981-01-01

A diagnostic system of the nuclear power plant operation is proposed and the applicability of this system to the actual plant has been verified by computer simulation. A typical pressurized water reactor plant simulator was made by an analog computer and the diagnostic system was made by a digital computer. The observed signals obtained from the actual plant are simulated by superposing the equivalent observation noises generated by the digital computer on the sampled signals obtained from the plant simulator. 8 refs

Training device for nuclear power plant operators

International Nuclear Information System (INIS)

Schoessow, G. J.

1985-01-01

A simulated nuclear energy power plant system with visible internal working components comprising a reactor adapted to contain a liquid with heating elements submerged in the liquid and capable of heating the liquid to an elevated temperature, a steam generator containing water and a heat exchanger means to receive the liquid at an elevated temperature, transform the water to steam, and return the spent liquid to the reactor; a steam turbine receiving high energy steam to drive the turbine and discharging low energy steam to a condenser where the low energy steam is condensed to water which is returned to the steam generator; an electric generator driven by the turbine; indicating means to identify the physical status of the reactor and its contents; and manual and automatic controls to selectively establish normal or abnormal operating conditions in the reactor, steam generator, pressurizer, turbine, electric generator, condenser, and pumps; and to be selectively adjusted to bring the reactor to acceptable operating condition after being placed in an abnormal operation. This device is particularly useful as an education device in demonstrating nuclear reactor operations and in training operating personnel for nuclear reactor systems and also as a device for conducting research on various safety systems to improve the safety of nuclear power plants

Development of a smart type motor operated valve for nuclear power plants

Science.gov (United States)

Kim, Chang-Hwoi; Park, Joo-Hyun; Lee, Dong-young; Koo, In-Soo

2005-12-01

In this paper, the design concept of the smart type motor operator valve for nuclear power plant was described. The development objective of the smart valve is to achieve superior accuracy, long-term reliability, and ease of use. In this reasons, developed smart valve has fieldbus communication such as deviceNet and Profibus-DP, auto-tuning PID controller, self-diagnostics, and on-line calibration capabilities. And also, to achieve pressure, temperature, and flow control with internal PID controller, the pressure sensor and transmitter were included in this valve. And, temperature and flow signal acquisition port was prepared. The developed smart valve will be performed equipment qualification test such as environment, EMI/EMC, and vibration in Korea Test Lab. And, the valve performance is tested in a test loop which is located in Seoul National University Lab. To apply nuclear power plant, the software is being developed according to software life cycle. The developed software is verified by independent software V and V team. It is expected that the smart valve can be applied to an existing NPPs for replacing or to a new nuclear power plants. The design and fabrication of smart valve is now being processed.

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

2005-01-01

This 2005 edition of the Elecnuc booklet summarizes in tables all numerical data relative to the nuclear power plants worldwide. These data come from the PRIS database managed by the IAEA. The following aspects are reviewed: 2004 highlights; main characteristics of reactor types; map of the French nuclear power plants on 2005/01/01; worldwide status of nuclear power plants at the end of 2004; units distributed by countries; nuclear power plants connected to the grid by reactor-type group; nuclear power plants under construction on 2004; evolution of nuclear power plant capacities connected to the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear power plants by country at the end 2004; performance indicator of PWR units in France; trend of the generation indicator worldwide; 2004 load factor by owners; units connected to the grid by countries at 12/31/2004; status of licence renewal applications in USA; nuclear power plants under construction at 12/31/2004; shutdown reactors; exported nuclear capacity in net MWe; exported and national nuclear capacity connected to the grid; exported nuclear power plants under construction or order; exported and national nuclear capacity under construction or order; recycling of plutonium in LWR; Mox licence plant projects; Appendix - historical development; acronyms, glossary

Nuclear power plant operator licensing

International Nuclear Information System (INIS)

1997-01-01

The guide applies to the nuclear power plant operator licensing procedure referred to the section 128 of the Finnish Nuclear Energy Degree. The licensing procedure applies to shift supervisors and those operators of the shift teams of nuclear power plant units who manipulate the controls of nuclear power plants systems in the main control room. The qualification requirements presented in the guide also apply to nuclear safety engineers who work in the main control room and provide support to the shift supervisors, operation engineers who are the immediate superiors of shift supervisors, heads of the operational planning units and simulator instructors. The operator licensing procedure for other nuclear facilities are decided case by case. The requirements for the basic education, work experience and the initial, refresher and complementary training of nuclear power plant operating personnel are presented in the YVL guide 1.7. (2 refs.)

Overview of the Italian activities in the field of life management of nuclear power plants

International Nuclear Information System (INIS)

Pini, A.

1994-01-01

Overview of the Italian activities in the field of life management of nuclear power plants is presented, including the following issues: nuclear energy scenario in Italy, study of reactor pressure vessel, piping and containment integrity, seismic risk analysis

Organizing nuclear power plant operation

International Nuclear Information System (INIS)

Adams, H.W.; Rekittke, K.

1987-01-01

With the preliminary culmination in the convoy plants of the high standard of engineered safeguards in German nuclear power plants developed over the past twenty years, the interest of operators has now increasingly turned to problems which had not been in the focus of attention before. One of these problems is the organization of nuclear power plant operation. In order to enlarge the basis of knowledge, which is documented also in the rules published by the Kerntechnischer Ausschuss (Nuclear Technology Committee), the German Federal Minister of the Interior has commissioned a study of the organizational structures of nuclear power plants. The findings of that study are covered in the article. Two representative nuclear power plants in the Federal Republic of Germany were selected for the study, one of them a single-unit plant run by an independent operating company in the form of a private company under German law (GmbH), the other a dual-unit plant operated as a dependent unit of a utility. The two enterprises have different structures of organization. (orig.) [de

Nuclear Power Plants in the World

International Nuclear Information System (INIS)

2003-01-01

The Japan Atomic Industrial Forum (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2002 was made up on bases of answers on questionnaires from 65 electric power companies and other nuclear organizations in 28 countries and regions around the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities; and so forth. (J.P.N.)

Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3. Behaviour of high pressure coolant injection system (HPCI) based on thermodynamic model

International Nuclear Information System (INIS)

Maruyama, Shigenao

2014-01-01

In order to clarify the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 3 is analyzed from the data open to the public. Phase equilibrium process model was introduced in which the vapor and water are at saturation point in the vessels. The present accident scenario assumes that the high pressure coolant injection system (HPCI) did not worked properly, but the steam in the reactor pressure vessel (RPV) leaked through the turbine of HPCI to the suppression chamber since 12/3/2011 12:35. It is assumed that the Tsunami flooded the torus room where the suppression chamber was placed. Proposed accident scenario agrees with the data of the plant parameters obtained just after the accident. It is estimated that the water injection by HPIC was stopped since around at 13/3 19:00 and the water level in RPV decreased since then. It is estimated that the RPV broke at 14/3 8:55 and water could injected from fire engines due to the depression due to the rupture of RPV. There was little water left in RPV at the time of the rupture. If the present scenario is correct, the behavior that operators in the plant stopped HPCI at 13/3 2:42 did not affect seriously on the RPV rupture. If HPCI was working properly until the operators stopped it, the plant parameters obtained in the accident cannot be explained. (author)

Inspection during operation of a nuclear power plant in Spain

International Nuclear Information System (INIS)

Gutierrez Bernal, R.

1977-01-01

The control and surveillance activities, as well as the operating data and results of the three nuclear power plants presently in operation: Jose Cabrera, Santa Maria de Garona and Vandellos, are summarized. The first two are light-water type, with different pressure and boiling characteristics and the third is of the gas-graphite type. The main aspects, from an inspection point of view, of the experience obtained in these three plants are analyzed. (author) [es

Large screen mimic display design research for advanced main control room in nuclear power plant

International Nuclear Information System (INIS)

Zheng Mingguang; Yang Yanhua; Xu Jijun; Zhang Qinshun; Ning Zhonghe

2002-01-01

Firstly the evolution of mimic diagrams or displays used in the main control room of nuclear power plant was introduced. The active functions of mimic diagrams were analyzed on the release of operator psychological burden and pressure, the assistance of operator for the information searching, status understanding, manual actuation, correct decision making as well as the safe and reliable operation of the nuclear power plant. The importance and necessity to use the (large screen) mimic diagrams in advanced main control room of nuclear power plant, the design principle, design details and verification measures of large screen mimic display are also described

Nuclear plant analyzer program for Bulgaria

International Nuclear Information System (INIS)

Shier, W.; Kennett, R.

1993-01-01

An interactive nuclear plant analyzer(NPA) has been developed for use by the Bulgarian technical community in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. The current NPA includes models for a VVER-440 Model 230 and a VVER-1000 Model 320 and is operational on an IBM RISC6000 workstation. The RELAP5/MOD2 computer code has been used for the calculation of the reactor responses to the interactive commands initiated by the NPA operator. The interactive capabilities of the NPA have been developed to provide considerable flexibility in the plant actions that can be initiated by the operator. The current capabilities for both the VVER-440 and VVER-1000 models include: (1) scram initiation; (2) reactor coolant pump trip; (3) high pressure safety injection system initiation; (4) low pressure safety injection system initiation; (5) pressurizer safety valve opening; (6) steam generator relief/safety valve opening; (7) feedwater system initiation and trip; (8) turbine trip; and (9) emergency feedwater initiation. The NPA has the capability to display the results of the simulations in various forms that are determined by the model developer. Results displayed on the reactor mask are shown through the user defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperature of other metal structures. In addition, changes in the status of various components and systems can be initiated and/or displayed both numerically and graphically on the mask. This paper provides a description of the structure of the NPA, a discussion of the simulation models used for the VVER-440 and the VVER-1000, and an overview of the NPA capabilities. Typical results obtained using both simulation models will be discussed

On nuclear power plant uprating

International Nuclear Information System (INIS)

Ho, S. Allen; Bailey, James V.; Maginnis, Stephen T.

2004-01-01

Power uprating for commercial nuclear power plants has become increasingly attractive because of pragmatic reasons. It provides quick return on investment and competitive financial benefits, while involving low risks regarding plant safety and public objection. This paper briefly discussed nuclear plant uprating guidelines, scope for design basis analysis and engineering evaluation, and presented the Salem nuclear power plant uprating study for illustration purposes. A cost and benefit evaluation of the Salem power uprating was also included. (author)

Simulation of a Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant. Simulacao do sistema nuclear de geracao de vapor de uma central PWR

Energy Technology Data Exchange (ETDEWEB)

Reis Martins Junior, L.L. dos.

1980-01-01

The following work intends to perform the digital simulation, of the Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant for control systems design and analysis purposes. There are mathematical models for the reactor, the steam generator, the pressurizer and for transport lags of the coolant in the primary circuit. Nevertheless no one control system has been considered to permit any user the inclusion in the more convenient way of the desired control systems' models. The characteristics of the system in consideration are fundamentally equal to the ones of Almirante Alvaro Alberto Nuclear Power Plant, Unit I (Angra I) obtained in the Final Safety Analysis Report at Comissao Nacional de Energia Nuclear. (author).

Nuclear power plant control and instrumentation 1982. Proceedings of an international symposium on nuclear power plant control and instrumentation

International Nuclear Information System (INIS)

1983-01-01

Ever increasing demands for nuclear power plant safety and availability imply a need for the introduction of modern measurement and control methods, together with data processing techniques based on the latest advances in electronic components, transducers and computers. Nuclear power plant control and instrumentation is therefore an extremely rapidly developing field. The present symposium, held in Munich, FR Germany, was prepared with the help of the IAEA International Working Group on Nuclear Power Plant Control and Instrumentation and organized in close co-operation with the Gesellschaft fur Reaktorsicherheit, Federal Republic of Germany. A number of developments were highlighted at the Munich symposium: - The increased use of computers can bring clear advantages and this technique is now proven as a tool for supervising and controlling plant operation. Advanced computerized systems for operator support are being developed on a large scale in many countries. The progress in this field is quite obvious, especially in disturbance analysis, safety parameter display, plant operator guidance and plant diagnostics. The new trend of introducing computers and microprocessors in protection systems makes it easy to implement 'defence-in-depth' strategies which give better assurance of correct system responses and also prevent unnecessary reactor trips, thus improving plant availability. The introduction of computerized systems for control of reactor power, reactor water level and reactor pressure as well as for reactor start-up and shut-down could improve the reliability and availability of nuclear power plants. The rapid technical development in the area of control and instrumentation makes it necessary to plan for at least one replacement of obsolete equipment in the course of the 30 years lifetime of a nuclear power plant and retrofitting of currently operating reactors with new control systems. Major design improvements and regulatory requirements also require

Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

International Nuclear Information System (INIS)

Bambang Galung Susanto

2007-01-01

The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

Effect of preemptive weld overlay on residual stress mitigation for dissimilar metal weld of nuclear power plant pressurizer

International Nuclear Information System (INIS)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae; Lee, Kyoung Soo; Park, Chi Yong

2008-01-01

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a Preemptive Weld OverLay(PWOL). In Pressurized Water Reactor(PWR) dissimilar metal weld is susceptible region for Primary Water Stress Corrosion Cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment

Effect of preemptive weld overlay on residual stress mitigation for dissimilar metal weld of nuclear power plant pressurizer

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-10-15

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a Preemptive Weld OverLay(PWOL). In Pressurized Water Reactor(PWR) dissimilar metal weld is susceptible region for Primary Water Stress Corrosion Cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

Nuclear security - New challenge to the safety of nuclear power plants

International Nuclear Information System (INIS)

Li Ganjie

2008-01-01

The safety of nuclear power plants involves two aspects: one is to prevent nuclear accidents resulted from systems and equipments failure or human errors; the other is to refrain nuclear accidents from external intended attack. From this point of view, nuclear security is an organic part of the nuclear safety of power plants since they have basically the same goals and concrete measures with each other. In order to prevent malicious attacks; the concept of physical protection of nuclear facilities has been put forward. In many years, a series of codes and regulations as well as technical standard systems on physical protection had been developed at international level. The United Nations passed No. 1540 resolution as well as 'Convention on the Suppression of Acts of Nuclear terrorism', and revised 'Convention on Physical Protection of Nuclear Materials', which has enhanced a higher level capacity of preparedness by international community to deal with security issues of nuclear facilities. In China, in order to improve the capability of nuclear power plants on preventing and suppressing the external attacks, the Chinese government consecutively developed the related codes and standards as well as technical documents based on the existing laws and regulations, including 'Guide for the Nuclear Security of Nuclear Power Plants' and 'Guide for the Physical Protection of Nuclear Materials', so as to upgrade the legislative requirements for nuclear security in power plants. The government also made greater efforts to support the scientific research and staff training on physical protection, and satisfying the physical protection standards for newly-built nuclear facilities such as large scale nuclear power plants to meet requirement at international level. At the same time old facilities were renovated and the Chinese government established a nuclear emergency preparedness coordination mechanism, developed corresponding emergency preparedness plans, intensified the

Nuclear power plants in the world

International Nuclear Information System (INIS)

2008-01-01

The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2007/2008 was made up on bases of answers on questionnaires from electric power companies and other nuclear organizations around the world by JAIF. This report is comprised of 18 items, and contains generating capacity of the plants; effect of the Niigata-ken chuetsu-oki earthquake; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities, and so forth. (J.P.N.)

Nuclear Power Plants in the World

International Nuclear Information System (INIS)

2004-01-01

The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2003 was made up on bases of answers on questionnaires from 81 electric power companies and other nuclear organizations in 33 countries and regions around the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities; and so forth. (J.P.N.)

Cooling facility of nuclear power plant

International Nuclear Information System (INIS)

Arai, Kenji; Nagasaki, Hideo.

1992-01-01

In a cooling device of a nuclear power plant, an exhaust pipe for an incondensible gas is branched. One of the branched exhaust pipes is opened in a pressure suppression pool water in a suppression chamber containing pool water and the other is opened at a lower portion of a dry well incorporating a pressure vessel. In a state where the pressure in the dry well is higher than that in the suppression chamber, an off-gas is exhausted effectively by way of the exhaustion pipe in communication with the suppression chamber. In a state where there is no difference between the pressures and the opening end of the exhaustion pipe in communication with the suppression chamber is sealed with water, off-gas is exhausted by way of the exhaustion pipe in communication with the lower portion of the dry well. Then, since the incondensible gas in a heat transfer pipe is not accumulated, after-heat can be removed efficiently. Satisfactory cooling is maintained even after the coincidence of the pressures in the dry well with that in the suppression chamber, to decrease a pressure in a reactor container. (N.H.)

Fatigue evaluation for the socket weld in nuclear power plants

International Nuclear Information System (INIS)

Choi, Young Hwan; Choi, Sun Yeong; Huh, Nam Soo

2004-01-01

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. Requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. The reduction of the weld leg size from 1.09 * t 1 to 0.75 * t 1 can affect severely on the socket weld integrity

Feasibility analysis of aggressive cooldown in OPR-1000 nuclear power plants

International Nuclear Information System (INIS)

Kim, Man Cheol; Jerng, Dong Wook

2014-01-01

Highlights: • We examine the feasibility of aggressive cooldown, particularly in OPR-1000 plants. • We review experimental and simulation results, EOPs, and time estimates. • Aggressive cooldown is, in general, expected to be performed successfully. • There is potential for failure to initiate aggressive cooldown in a timely manner. • Investigation on allowing higher cooldown rate in OPR-1000 plants is recommended. - Abstract: Aggressive cooldown is the action taken by main control room (MCR) operators to rapidly cool down and depressurize the reactor coolant system at the maximum allowed cooldown rate during a small break loss of coolant accident (SBLOCA) or a steam generator tube rupture accident with failure of high-pressure safety injection (HPSI) in a pressurized water reactor. We examined the feasibility of aggressive cooldown in OPR-1000 nuclear power plants by establishing success criteria based on experimental results and thermal–hydraulic code simulations, reviewing emergency operating procedures, and estimating the necessary time for MCR operators to initiate aggressive cooldown based on experimental results with licensed nuclear power plant operators. For an SBLOCA with a break size of 0.02 ft 2 combined with total failure of HPSI, we found that aggressive cooldown can generally be performed successfully, but there is potential for failure to initiate aggressive cooldown in a timely manner. We discuss the potential effects of introducing the Ultimate Procedure on reducing the core damage potential. Detailed analysis and further discussion are necessary to reduce uncertainties associated with aggressive cooldown and its positive effect on the safety of nuclear power plants

Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

International Nuclear Information System (INIS)

Kohler, H.A.G.

1977-01-01

This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

Political pressure on nuclear - responsibility or business?

International Nuclear Information System (INIS)

Petrech, Rastislav; Holy, Robert

2001-01-01

respectively). Without this compromise of the Slovak Government we would not be invited to talks on joining the European Union. There is a similar situation in Bulgaria and Lithuania. The governments of the countries were forced to a compromise solutions and promised that Kozloduy or Ignalina NPP's respectively would be shut down in a close future, to be invited to talks about joining the EU. Another example is Sweden. Based on the pressure of greens to shut down NPP's, the unit 1 of Barsebaeck was shut down (in 1999 - with no technical justification) for a huge governmental compensation to the operator. This resulted in increased electricity imports from Danish and German coal-fired plants - causing indirect rise in CO 2 emission by 4 million tons per year (8 % of the total Swedish emissions). This is probably why 78% of the Swedish disagree with nuclear power phaseout. It is confirmed that the plans of premature shutdown of Barsebaeck-2 is postponed from the originally suggested date (1st July 2001). In Germany also more than 60 % of the population believes that step-by-step shut down of NPP's is not realistic in short-term. There are indices from other countries with well-developed nuclear power industry, such as USA, France, and Finland, that nuclear power renaissance can be expected in the future. The fresh examples of the Czech Temelin NPP and Mochovce (Slovakia) are very similar: halt of construction due to financial reasons, replacement of some plant systems, completion under a strong opposition and political pressure of greens supported by Austrian Government, the same accusations and complaints, etc. The opponents also have similar scenarios in both cases. The Austrian and greens started massive campaigns prior to initial fuel loading and were trying very hard to postpone the commissioning process. Austria lost the nuclear war in Slovakia and so will they lose in the Czech Republic. The compromise promised by the Czech Prime Minister - development of a new

The art and trend of nuclear power plants aging management and licenses renewal activity In USA

International Nuclear Information System (INIS)

Zhong Zhimin; Li Jinsong; Zhang Mengyi

2014-01-01

This paper briefly introduced the history and the art of nuclear power plants licenses renewal in United State. The aims, working scope, methodology, the art and trend of aging management and its role in license renewal process in United State nuclear power plants license renewal process were discussed in details. Furthermore, the aging management current research focus in United State was described. Then, take into account the AP serials Pressurized Water Reactor and nuclear safety requirements in the regulatory and safety guide in China, some suggestions and recommendation on nuclear power plants aging management were introduced, which will be helpful when we developed related aging management works in China. (authors)

The fourth nuclear power plant in Argentina

Energy Technology Data Exchange (ETDEWEB)

Nogarin, Mauro

2017-01-15

Since 2006 the nuclear sector in Argentina has aimed at recovering and strengthening its capabilities and facilities. Part of the challenge posed by this revival has been to also accompany the development of activities with a higher level of responsibility for safety and the environment. Among the strategic decisions taken in recent years, one main highlight is the construction of the nuclear power plant CAREM25 entirely with Argentine technology and design under the responsibility of the National Atomic Energy Commission. On February 4, 2015, the Ministry of Federal Planning and the National Energy Administration (NEA) signed the agreement for cooperation and construction of pressurized water reactor (PWR) with ACP-1000 technology, developed in the Peoples Republic of China.

Near term feasibility of nuclear reactor for sea-water desalting: coupling of standard condensing nuclear power stations to low grade heat multieffect distillation plants

International Nuclear Information System (INIS)

Adar, J.; Manor, S.; Schaal, M.

1977-01-01

Commercial nuclear power reactors exist only in standard sizes and designs. No large nuclear back-pressure turbines are available today. Therefore, near term large scale nuclear desalination plants must be tailored to the NSSS sizes and available turbines and not the contrary. Standard condensing nuclear turbines could operate continuously with a back-presure of up to 5-7'' Hg (depending on the supplier). It means that they can exhaust huge amounts of steam at 56 0 C - 64 0 C with a loss of electricity production of 6% - 10% when compared to 2 1/2'' Hg normal condensing pressure. The horizontal aluminium tube multi-effect distillation process developed by ''Israel Desalination Engineering'' Ltd. is very suitable for the use of such low-grade heat: 4 to 9 effects can operate within these temperature ranges. A special flash-chamber constitutes a positive barrier against any possible contamination being carried over by the steam exhausted from the turbine to the desalination plant. Flow sheets, heat and mass balances have been prepared for two standard sizes of NSSS and turbines (1882sup(Mwth) and 2785sup(Mwth)), two ''back-pressures'' (5 1/2'' and 7'' Hg), and corresponding desalination plants. Only standard equipment is being used in the steam and electricity producing plant. The desalination plant consists of 6 to 12 parallel double lines, each of them similar to a large prototype now being designed and which is going to be coupled to an old fossil power station. Water production varies between 50 and 123 sup(us MGD) and water cost between 23 and 36 sup(cents)/M 3 . Total energy requirements of the desalination plant represent only 19 to 50% of the total water cost as against 75% for a single purpose plant. Costs are based on actual bids for the power plant and actual estimates for the desalination prototype. The operation is designed to be flexible so that the power plant can be operated either in conjunction with the desalination plant, or as a single purpose

Nuclear Power Plants (Rev.)

Energy Technology Data Exchange (ETDEWEB)

Lyerly, Ray L.; Mitchell III, Walter [Southern Nuclear Engineering, Inc.

1973-01-01

Projected energy requirements for the future suggest that we must employ atomic energy to generate electric power or face depletion of our fossil-fuel resources—coal, oil, and gas. In short, both conservation and economic considerations will require us to use nuclear energy to generate the electricity that supports our civilization. Until we reach the time when nuclear power plants are as common as fossil-fueled or hydroelectric plants, many people will wonder how the nuclear plants work, how much they cost, where they are located, and what kinds of reactors they use. The purpose of this booklet is to answer these questions. In doing so, it will consider only central station plants, which are those that provide electric power for established utility systems.

Improvement of the nuclear plant analyzer for Korean Standard Nuclear Power Plants

International Nuclear Information System (INIS)

Choi, Sung Soo; Han, Byoung Sub; Suh, Jae Seung; Son, Dae Seong

2005-04-01

Accurate analysis of the operating characteristics of Nuclear Power Plants provides valuable information for both norman and abnormal operation. The information can be used for the enhancement of plant performance and safety. Usually, such analysis is performed using computer codes used for plant design or simulators. However, their usages are limited because special expertise is required to use the computer codes and simulators are not portable. Therefore, it deemed necessary to develop an NPA which minimizes those limitations and can be used for the analysis and simulation of Nuclear Power Plants. The purpose of this study is to develop a real-time best-estimate NPA for the Korean Nuclear Power Plants(KSNP). The NPA is an interactive, high fidelity engineering simulator. NPA combines the process model simulating the plant behavior with the latest computer technology such as Graphical User Interface(GUI) and simulation executive for enhanced user interface. The process model includes models for a three-dimensional reactor core, the NSSS, secondary system including turbine and feedtrain, safety auxiliary systems, and various control systems. Through the verification and validation of the NPA, it was demonstrated that the NPA can realistically simulate the plant behaviors during transient and accident conditions

Nuclear power plants in post-war thought

International Nuclear Information System (INIS)

Toya, Hiroshi

2015-01-01

This paper overviews how nuclear power plants have been talked about in the post-war thought. Science and technology sometimes significantly change the thinking way of humans, and nuclear power generation is an extreme technology. This paper overviews how nuclear power plants and humans are correlated. The following three points are discussed as the major issues of contemporary thought over nuclear power plants. First, on the danger of nuclear power plants, the risk of destructive power that nuclear energy has, and the danger of unreasoning development in science and technology civilization are discussed. Second, on the ethics issues surrounding nuclear power plants, the ethics that are based on unbalanced power relations, and democratic responsibility ethics based on discussion ethics are discussed. Third, on the issues of nuclear power plants and imagination, the limitations of democratic discussion surrounding nuclear power plants, the formation of imagination commensurate with the destructive power of nuclear power plants, and the formation of imagination that can represent the distant future are discussed. (A.O.)

Public regulation of nuclear plants

International Nuclear Information System (INIS)

Burtheret, M.; Cormis, de

1980-01-01

The construction and operation of nuclear plants are subject to a complex system of governmental administration. The authors list the various governmental authorisations and rules applicable to these plants. In the first part, they describe the national regulations which relate specifically to nuclear plants, and emphasize the provisions which are intended to ensure the safety of the installations and the protection of the public against ionizing radiation. However, while the safety of nuclear plants is a major concern of the authorities, other interests are also protected. This is accomplished by various laws or regulations which apply to nuclear plants as well as other industrial installations. The duties which these texts, and the administrative practice based thereon, impose on Electricite de France are covered in the second part [fr

Analysis of failed nuclear plant components

Science.gov (United States)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

Experience with diagnostic instrumentation in nuclear power plants

International Nuclear Information System (INIS)

Gopal, R.; Ciaramitaro, W.

1977-01-01

Over the past several years, Westinghouse has developed a coordinated system of on-line diagnostic instrumentation for the acquisition and analysis of data for diagnostics and incipient failure detection of critical plant equipment and systems. Primary motivation for this work is to improve NSSS availability and Maintainability through the detection of malfunctions at their inception. These systems encompass the following areas: (1) Vibration Monitoring System for detection of changes in vibrational characteristics of the major components of Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP); (2) Acoustic Monitoring System for detection and location of leaks in the primary system pressure boundary and other piping systems in PWRs; (3) Metal Impact Monitoring for detection of loose debris in the reactor vessel and steam generators; (4) Nuclear Noise Monitoring System for monitoring core barrel vibration; (5) Sensor Response Time Measurement System for detecting any degradation of process sensors; and (6) Transit Time Flow Meter for determining primary coolant flow rate. Summarized in this paper are some of the features of the systems and in-plant experience. These experiences demonstrate that diagnostic systems in combination with analytical and laboratory work for data interpretation do improve plant availability. (author)

Performances of nuclear power plants for combined production of electricity and hot water for district heating

International Nuclear Information System (INIS)

Bronzen, S.

The possibilities for using nuclear power plants for combined production of heat and power seem to be very good in the future. With the chosen 600 MWsub (e) BWR plant a heat output up to 1200 MW can be arranged. An alternative, consisting of steam extractions from the low-pressure turbine, offers a flexible solution for heat and power generation. With this alternative the combined plant can use components from normal condensing nuclear power plants. The flexible extraction design also offers a real possibility for using the combined plant in electric peak generation. However, urban siting requires long distance heat transmission and the pipe design for this transmission is a major problem when planning and optimizing the whole nuclear combined heat and power plant. (author)

Nuclear Power Plants in the World

International Nuclear Information System (INIS)

2000-01-01

The Japan Atomic Industrial Forum (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Developmental trends on nuclear power plants in the world'. In this report, some data at the end of 1999 was made up on bases of answers on questionnaires from 72 electric companies in 31 nations and regions in the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; location of the plants; the plants in the world; and so forth. And, it also has some survey results on the 'Liberalization of electric power markets and nuclear power generation' such as some 70% of respondents in nuclear power for future option, gas-thermal power seen as power source with most to gain from liberalization, merits on nuclear power generation (environmental considerations and supply stability), most commonly voiced concern about new plant orders in poor economy, and so forth. (G.K.)

Erosion/corrosion-induced pipe wall thinning in US Nuclear Power Plants

International Nuclear Information System (INIS)

Wu, P.C.

1989-04-01

Erosion/corrosion in single-phase piping systems was not clearly recognized as a potential safety issue before the pipe rupture incident at the Surry Power Station in December 1986. This incident reminded the nuclear industry and the regulators that neither the US Nuclear Regulatory Commission (NRC) nor Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code require utilities to monitor erosion/corrosion in the secondary systems of nuclear power plants. This report provides a brief review of the erosion/corrosion phenomenon and its major occurrence in nuclear power plants. In addition, efforts by the NRC, the industry, and the ASME Section XI Committee to address this issue are described. Finally, results of the survey and plant audits conducted by the NRC to assess the extent of erosion/corrosion-induced piping degradation and the status of program implementation regarding erosion/corrosion monitoring are discussed. This report will support a staff recommendation for an additional regulatory requirement concerning erosion/corrosion monitoring. 21 refs., 3 tabs

Gland system, especially for nuclear power plant circulation pumps

International Nuclear Information System (INIS)

Skalicky, A.; Vesely, M.

1975-01-01

The invention claims a gland system suitable especially for the circulation pumps of nuclear power plants. The system prevents the release of the radioactive high-pressure cooling liquid in the atmosphere. The gland system consists of at least two mechanical glands arranged in series and of the closed circuit of the cooling high-pressure medium. The respective mechanical glands are linked with by-pass branches and discharge piping. The by-pass branches accommodating control manometers and flowmeters are linked with the storage reservoir with drain pipes provided with stop fittings. (Oy)

Influence of Egyptian electrical grid and nuclear power plants under disturbances based on PSS/E

International Nuclear Information System (INIS)

Shaat, M. K.; Kotb, S. A.; Mahmoud, H. M.

2012-12-01

The capacity of the electrical power system in Egypt will increase rapidly in the coming twenty years. In year 2018, power generation will be connecting to the Egyptian electrical grid. Consequently, the interaction of nuclear power plants and other systems become a very important issue, and a detailed nuclear power model for the medium-term and long-term power system stability should be developed. However, there is no nuclear unit model that can describe the detailed characteristics of the nuclear unit in the available commercial power system simulation software. In this paper, a detailed pressurized water reactor (PWR) nuclear unit model for medium-term and long-term power system transient stability is proposed. The model is implemented by a user defined program in PSS/E through PSS/E Mat lab Seamanlike interface. Also this paper proposes a design of power plant rector controller for the nuclear power plant. This model can be used to analyze the difference influences between the Egyptian electrical grid and nuclear power plants for examples transient fault on electrical grid and outage of nuclear power plant. The simulation results show that the proposed model is valid. (Author)

A simulation-based expert system for nuclear power plant diagnostics

International Nuclear Information System (INIS)

Hassberger, J.A.; Lee, J.C.

1989-01-01

An expert system for diagnosing operational transients in a nuclear power plant is discussed. Hypothesis and test is used as the problem-solving strategy with hypotheses generated by an expert system that monitors the plant for patterns of data symptomatic of known failure modes. Fuzzy logic is employed as the inferencing mechanism with two complementary implication schemes to handle scenarios involving competing failures. Hypothesis testing is performed. An artificial intelligence framework based on a critical functions approach is used to deal with the complexity of a nuclear plant. A prototype system for diagnosing transients in the reactor coolant system of a pressurized water reactor has been developed to test the algorithms described here. Results are presented for the diagnosis of data from the Three Mile Island Unit 2 loss-of-feedwater/small-break loss-of-collant accident

Effect of marine condition on feature of natural circulation after accident in floating nuclear power plant

International Nuclear Information System (INIS)

Yang Fan; Zhang Dan; Tan Changlu; Ran Xu; Yu Hongxing

2015-01-01

The incline and swing effect on natural circulation of floating nuclear power plant under site black out (SBO) accident is studied using self-developing marine condition system code RELAP5/MC. It shows that, for floating nuclear power plant under marine condition, the pressurizer fluctuating flow rate, the parallel heat sink (steam generator) have significant influences on the direct passive reactor heat removal (PRHR) system, which is different from other secondary PRHR under marine condition. The flow exchange between the loop and the pressurizer have major effect on cooling capacity for the left side loop. (authors)

Hygiene problems in building a nuclear power and heat plant near Bratislava

International Nuclear Information System (INIS)

Chorvat, D.; Mizov, J.; Hladky, E.; Kubik, I.; Carach, J.

1976-01-01

The results are presented of the calculation of the population exposure due to the release of radioactive products from a nuclear power and heating plant accident into the ambient atmosphere (primary coolant circuit rupture) providing the nuclear power and heating plant is sited approximately 500 m from the Slovnaft chemical works in Bratislava. Ground water contamination was analyzed assuming the infiltration of radioactive products from a surface deposit due to fallout and the direct infiltration of the products into the soil in the area of the plant. The results of the assessment of the design basis accident of a WWER-1000 nuclear power and heating plant show unequivocally that the emergency core cooling system, full-pressure containment and the correct function of the containment spray system are able to keep the accident consequences within acceptable limits thus meeting radiation hygiene requirements related to the siting of similar installations in the vicinity of large housing estates. (Oy)

Nuclear power plant safety

International Nuclear Information System (INIS)

Otway, H.J.

1974-01-01

Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

Seismic isolation of nuclear power plants - EDF's philosophy

International Nuclear Information System (INIS)

Coladant, C.

1989-01-01

The elastomer bearing pads used since 1963 as supports for prestressed concrete pressure vessels (PCPVs) was quickly chosen by Electricite de France (ED) to improve the capability of nuclear power plants (NPPs) to withstand strong earthquakes and to reduce the seismic loads on structures and equipment. The standardized units for 900 and 1,300 MW(e) pressurized water reactor (PWR) plants have moderate seismic design loads of 0.2 and 0.15 g, respectively. These design loads were exceeded by the site dependent spectra of Cruas (France) and Koeberg (South Africa). To keep the plant design unchanged and to take the advantages of standardization, these units were put on laminated bearings with or without sliding plates. For the future French 1,500 MW(e) fast breeder reactors (FBRs), which are more sensitive to seismic loads, the base isolation is considered by EDF at the beginning of the design, even for low ground motions of 0.1 g. The buildings are placed on laminated bearings while the reactor block is supported by springs and dampers. The isolated plant has identical costs as a conventional design such as SPX1 at Creys-Malville

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

1998-01-01

This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1997 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1997; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; forecasts; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

Coupling of standard condensing nuclear power stations to horizontal aluminium tubes multieffect distillation plants

International Nuclear Information System (INIS)

Adar, J.

1977-01-01

No large nuclear back-pressure turbines are available to-day. Standard condensing nuclear turbines could operate continuously with a back-pressure of up to 7'' Hg, exhausting huge amounts of steam at 56degC-64degC with a loss of electricity production of only 6%-10%. The horizontal aluminium tube multieffect distillation process developed by 'Israel Desalination Engineering Ltd' is very suitable for the use of such low-grade heat. A special flash-chamber loop constitutes a positive barrier against any possible contamination being carried over by the steam exhausted from the turbine to the desalination plant. The operation is designed to be flexible so that the power plant can be operated either in conjunction with the desalination plant or as a single purpose plant. Flow sheets, heat and mass balances have been prepared for eight different combinations of plants. Only standard equipment is being used in the power plant. The desalination plant consists of 6 to 12 parallel double lines, each of them similar to a large prototype now being designed. Water production varies between 50 and 123 MGD and water cost between 90 and 137 c/1000 gallons. Costs are based on actual bids

KWU Nuclear Plant Analyzer

International Nuclear Information System (INIS)

Bennewitz, F.; Hummel, R.; Oelmann, K.

1986-01-01

The KWU Nuclear Plant Analyzer is a real time engineering simulator based on the KWU computer programs used in plant transient analysis and licensing. The primary goal is to promote the understanding of the technical and physical processes of a nuclear power plant at an on-site training facility. Thus the KWU Nuclear Plant Analyzer is available with comparable low costs right at the time when technical questions or training needs arise. This has been achieved by (1) application of the transient code NLOOP; (2) unrestricted operator interaction including all simulator functions; (3) using the mainframe computer Control Data Cyber 176 in the KWU computing center; (4) four color graphic displays controlled by a dedicated graphic computer, no control room equipment; and (5) coupling of computers by telecommunication via telephone

Nuclear accidents and safety measures of domestic nuclear power plants

International Nuclear Information System (INIS)

Song Zurong; Che Shuwei; Pan Xiang

2012-01-01

Based on the design standards for the safety of nuclear and radiation in nuclear power plants, the three accidents in the history of nuclear power are analyzed. And the main factors for these accidents are found out, that is, human factors and unpredicted natural calamity. By combining the design and operation parameters of domestic nuclear plants, the same accidents are studied and some necessary preventive schemes are put forward. In the security operation technology of domestic nuclear power plants nowadays, accidents caused by human factors can by prevented completely. But the safety standards have to be reconsidered for the unpredicted neutral disasters. How to reduce the hazard of nuclear radiation and leakage to the level that can be accepted by the government and public when accidents occur under extreme conditions during construction and operation of nuclear power plants must be considered adequately. (authors)

Perspectives of nuclear power plants

International Nuclear Information System (INIS)

Vajda, Gy.

2001-01-01

In several countries the construction of nuclear power plants has been stopped, and in some counties several plants have been decommissioned or are planned to. Therefore, the question arises: have nuclear power plants any future? According to the author, the question should be reformulated: can mankind survive without nuclear power? To examine this challenge, the global power demand and its trends are analyzed. According to the results, traditional energy sources cannot be adequate to supply power. Therefore, a reconsideration of nuclear power should be imminent. The economic, environmental attractions are discussed as opposite to the lack of social support. (R.P.)

Dynamic analysis of Boushehr Nuclear Power Plant in connected to grid system

International Nuclear Information System (INIS)

Karimi Fard, A.

1999-01-01

Models of generating with the pressurized water reactor (PWR) have been developed for simulating. the plant dynamics under system disturbances. These models include power plant, energy sources, turbine, transmission system and control system such as Avr and govern and other local control devices. Simulink toolbox of Matlab software is used for simulations. The study is mainly based on the Bushehr Nuclear Power Plants (BNPP) parameters. Assuming that BNPP is connected to infinite bus with double tie line. Four cases are studied to examine the internal dynamic behavior of BNPP. First and second cases are used to load following studies in nuclear power plant. Another cases are used to study the dynamic behavior after short circuit fault and line outages in transmission systems. The results discussed in the thesis

technical guidelines for the design and construction of the next generation of nuclear power plants with pressurized water reactors

International Nuclear Information System (INIS)

2009-01-01

These technical guidelines present the opinion of the French 'Groupe Permanent charge des Reacteurs nucleaires' (GPR) concerning the safety philosophy and approach as well as the general safety requirements to be applied for the design and construction of the next generation of nuclear power plants of the PWR (pressurized water reactor) type, assuming the construction of the first units of this generation would start at the beginning of the 21. century. These technical guidelines are based on common work of the French Institut de Protection et de Surete Nucleaire (IPSN) and of the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Moreover, these technical guidelines were extensively discussed with members of the German Reaktor Sicherheitskommission (RSK) until the end of 1998 and further with German experts. The context of these technical guidelines must be clearly understood. Faced with the current situation of nuclear energy in the world, the various nuclear steam supply system designers are developing new products, all of them claiming their intention of obtaining a higher safety level, by various ways. GPR believes that, for the operation of a new series of nuclear power plants at the beginning of the next century, the adequate way is to derive the design of these plants in an 'evolutionary' way from the design of existing plants, taking into account the operating experience and the in-depth studies conducted for such plants. Nevertheless, introduction of innovative features must also be considered in the frame of the design of the new generation of plants, especially in preventing and mitigating severe accidents. GPR underlines here that a significant improvement of the safety of the next generation of nuclear power plants at the design stage is necessary, compared to existing plants. If the search for improvement is a permanent concern in the field of safety, the necessity of a significant step at the design stage clearly derives from better

Off-shore nuclear power plant

International Nuclear Information System (INIS)

Nakanishi, T.

1980-01-01

In order to avoid losses of energy and seawater pollution an off-shore nuclear power plant is coupled with a power plant which utilizes the temperature difference between seawater and hot reactor cooling water. According to the invention the power plant has a working media loop which is separated from the nuclear power plant. The apparative equipment and the operational characteristics of the power plant are the subject of the patent. (UWI) [de

Plant life management optimized utilization of existing nuclear power plants

International Nuclear Information System (INIS)

Watzinger, H.; Erve, M.

1999-01-01

For safe, reliable and economical nuclear power generation it is of central importance to understand, analyze and manage aging-related phenomena and to apply this information in the systematic utilization and as-necessary extension of the service life of components and systems. An operator's overall approach to aging and plant life management which also improves performance characteristics can help to optimize plant operating economy. In view of the deregulation of the power generation industry with its increased competition, nuclear power plants must today also increasingly provide for or maintain a high level of plant availability and low power generating costs. This is a difficult challenge even for the newest, most modern plants, and as plants age they can only remain competitive if a plant operator adopts a strategic approach which takes into account the various aging-related effects on a plant-wide basis. The significance of aging and plant life management for nuclear power plants becomes apparent when looking at their age: By the year 2000 roughly fifty of the world's 434 commercial nuclear power plants will have been in operation for thirty years or more. According to the International Atomic Energy Agency, as many as 110 plants will have reached the thirty-year service mark by the year 2005. In many countries human society does not push the construction of new nuclear power plants and presumably will not change mind within the next ten years. New construction licenses cannot be expected so that for economical and ecological reasons existing plants have to be operated unchallengeably. On the other hand the deregulation of the power production market is asking just now for analysis of plant life time to operate the plants at a high technical and economical level until new nuclear power plants can be licensed and constructed. (author)

Proceedings of the topical meeting on nuclear power plant life extension

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book contains the proceedings of the topical meeting on nuclear power plant life extension. The sessions are organized under the following headings: Perspectives on nuclear power plant life extension, the potential for additional years of power production, NRC and industry life extension initiatives, concrete and structures degradation and evaluation of useful remaining life, plant life extension programs, Reactor pressure vessel and intervals degradation and evaluation of useful remaining life, life extension decision making issues and institutions, systems degradation and evaluation of remaining life, monitoring and repair, design records and maintenance activities for life extension, Mechanical and electrical component degradation and evaluation of remaining life, expert systems and other techniques for enhanced and continued operation, life extension aspect of codes, standards, and related technologies, piping and valve degradation and evaluation of useful remaining life

Building of nuclear power plant

International Nuclear Information System (INIS)

Saito, Takashi.

1997-01-01

A first nuclear plant and a second nuclear power plant are disposed in adjacent with each other in a building for a nuclear reactor. A reactor container is disposed in each of the plants, and each reactor container is surrounded by a second containing facility. A repairing chamber capable of communicating with the secondary containing facilities for both of the secondary containing facilities is disposed being in contact with the second containing facility of each plant for repairing control rod driving mechanisms or reactor incorporated-type recycling pumps. Namely, the repairing chamber is in adjacent with the reactor containers of both plants, and situated between both of the plants as a repairing chamber to be used in common for both plants. Air tight inlet/exit doors are formed to the inlets/exits of both plants of the repairing chamber. Space for the repairing chamber can be reduced to about one half compared with a case where the repairing chamber is formed independently on each plant. (I.N.)

Neutron dosimetry at commercial nuclear plants. Final report of Subtask B: dosimeter response

Energy Technology Data Exchange (ETDEWEB)

Cummings, F.M.; Endres, G.W.R.; Brackenbush, L.W.

1983-03-01

As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterize neutron dosimeter responses inside the containment structure of commercial nuclear plants. In order to characterize those responses, dosimeters were irradiated inside containment at 2 pressurized water reactors and at pipe penetrations outside the biological shield at two boiling water reactors. The reactors were operating at full power during the irradiations. Measurements were also performed with electronic instruments, the tissue equivalent proportional counter (TEPC), and portable remmeters, SNOOPY, RASCAL and PNR-4.

Neutron dosimetry at commercial nuclear plants. Final report of Subtask B: dosimeter response

International Nuclear Information System (INIS)

Cummings, F.M.; Endres, G.W.R.; Brackenbush, L.W.

1983-03-01

As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterize neutron dosimeter responses inside the containment structure of commercial nuclear plants. In order to characterize those responses, dosimeters were irradiated inside containment at 2 pressurized water reactors and at pipe penetrations outside the biological shield at two boiling water reactors. The reactors were operating at full power during the irradiations. Measurements were also performed with electronic instruments, the tissue equivalent proportional counter (TEPC), and portable remmeters, SNOOPY, RASCAL and PNR-4

Nuclear Power Plants. Revised.

Science.gov (United States)

Lyerly, Ray L.; Mitchell, Walter, III

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

Service life monitoring of the main components at the Temelin nuclear power plant

International Nuclear Information System (INIS)

Hahn, J.; Vincour, D.

2007-01-01

Knowledge and experience gained from the introduction and periodical implementation of life assessment of the major components of the Temelin nuclear power plant is summarized. The initial Soviet technical design of the plant did not incorporate lifetime monitoring and evaluation, therefore it was completed with demonstrative strength and lifetime calculations from Czech companies. Moreover, a Westinghouse primary circuit diagnosis and monitoring system, including the monitoring of temperature and pressure cycles for low-cycle fatigue evaluation, was installed at the plant. The DIALIFE code for the calculation of mainly the low-cycle fatigue of the key pressure components, was developed and installed subsequently as a superstructure to the monitoring system. (author)

Development of technologies on innovative-simplified nuclear power plant using high-efficiency steam injectors. (2) Analysis of heat balance of innovative-simplified nuclear power plant

International Nuclear Information System (INIS)

Goto, Shoji; Ohmori, Shuichi; Mori, Mitchitsugu

2004-01-01

It is possible to established simplified systems and reduced space and equipments using high-efficiency Steam Injector (SI) instead of low-pressure feed water heaters in Nuclear Power Plant (NPP). The SI works as a heat exchanger through direct contact between feedwater from condenser and extracted steam from turbine. It can get a higher pressure than supplied steam pressure, so it can reduce the feedwater pumps. The maintenance and reliability are still higher because SI has no movable parts. This paper describes the analysis of the heat balance and plant efficiency of this Innovative-Simplified NPP with high-efficiency SI. The plant efficiency is compared with the electric power of 1100MWe class original BWR system and the Innovative-Simplified BWR system with SI. The SI model is adapted into the heat balance simulator with a simplified model. The results show plant efficiencies of the Innovated-Simplified BWR system are almost equal to the original BWR one. The present research is one of the projects that are carried out by Tokyo Electric Power Company, Toshiba Corporation, and six Universities in Japan, funded from the Institute of Applied Energy (IAE) of Japan as the national public research-funded program. (author)

75 FR 66802 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

Science.gov (United States)

2010-10-29

... Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2; Notice of Withdrawal of...) has granted the request of Calvert Cliffs Nuclear Power Plant, LLC, the licensee, to withdraw its... for the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, located in Calvert County, MD. The...

Development of construction technology for nuclear power plants

International Nuclear Information System (INIS)

Schnellenbach, G.

1975-01-01

Stringent safety requirements for nuclear power plants create new technological problems in their construction. The development is influenced by the mode of operation of the respective reactors which leads to different construction styles of the reactor buildings. Accomodation of extraordinary load cases such as earthquake, airplane crash, and blast due to chemical explosions, requires additional treatment. Significant new problems arise for the prestressed concrete reactor pressure vessels and concrete containments, prestressed or reinforced. (orig.) [de

Electromagnetic compatibility of nuclear power plants

International Nuclear Information System (INIS)

Cabayan, H.S.

1983-01-01

Lately, there has been a mounting concern about the electromagnetic compatibility of nuclear-power-plant systems mainly because of the effects due to the nuclear electromagnetic pulse, and also because of the introduction of more-sophisticated and, therefore, more-susceptible solid-state devices into the plants. Questions have been raised about the adequacy of solid-state-device protection against plant electromagnetic-interference sources and transients due to the nuclear electromagnetic pulse. In this paper, the author briefly reviews the environment, and the coupling, susceptibility, and vulnerability assessment issues of commercial nuclear power plants

Nuclear power plants in populated areas

International Nuclear Information System (INIS)

Wachsmann, F.

1973-01-01

The article first deals with the permanently increasing demand for electical power. Considering the ever growing energy demand which can no longer be covered by conventional power plants, it has become necessary to set up nuclear power plants of larger range. The author presents in a survey the basic function of nuclear power plants as well as the resulting risks and safety measures. The author concludes that according to present knowledge there is no more need to erect nuclear power plants outside densely populated urban areas but there is now the possibility of erecting nuclear power plants in densely populated areas. (orig./LH) [de

Radiological Effluent Technical Specifications (RETS) implementation, Kewaunee Nuclear Power Plant

International Nuclear Information System (INIS)

Serrano, W.; Akers, D.W.

1985-06-01

A review of the Radiological Effluent Technical Specifications (RETS) for the Kewaunee Nuclear Power Plant was performed. The principal review guidelines used were NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,'' and Draft 7'' of NUREG-0472, Revision 3, ''Radiological Effluent Technical Specifications for Pressurized Water Reactors.'' Draft submittals were discussed with the Licensee by the NRC staff until all items requiring changes to the Technical Specifications were resolved. The Licensee then submitted final proposed RETS to the NRC which were evaluated and found to be in compliance with the NRC review guidelines. The proposed Offsite Dose Calculation Manual and the Radiological Environmental Monitoring Manual were reviewed and generally found to be in compliance with the NRC review guidelines

Advanced digital technology - improving nuclear power plant performance through maintainability

International Nuclear Information System (INIS)

Ford, J.L.; Senechal, R.R.; Altenhein, G.D.; Harvey, R.P.

1998-01-01

In today's energy sector there is ever increasing pressure on utilities to operate power plants at high capacity factors. To ensure nuclear power is competitive into the next century, it is imperative that strategic design improvements be made to enhance the performance of nuclear power plants. There are a number of factors that affect a nuclear power plant's performance; lifetime maintenance is one of the major contributors. The maturing of digital technology has afforded ABB the opportunity to make significant design improvements in the area of maintainability. In keeping with ABB's evolutionary advanced nuclear plant design approach, digital technology has systematically been incorporated into the control and protection systems of the most recent Korean nuclear units in operation and under construction. One example of this was the multi-functional design team approach that was utilized for the development of ABB's Digital Plant Protection System (DPPS). The design team consisted of engineers, maintenance technicians, procurement specialists and manufacturing personnel in order to provide a complete perspective on all facets of the design. The governing design goals of increased reliability and safety, simplicity of design, use of off-the-shelf products and reduced need for periodic surveillance testing were met with the selection of proven ABB-Advant Programmable Logic Controllers (PLCs) as the heart of the DPPS. The application of digital PLC technology allows operation for extended periods without requiring routine maintenance or re-calibration. A well documented commercial dedication program approved by the United States Nuclear Regulatory Commission (US NRC) as part of the System 80+ TM Advanced Light Water Reactor Design Certification Program, allowed the use of off-the shelf products in the design of the safety protection system. In addition, a number of mechanical and electrical improvements were made which support maintainability. The result is a DPPS

Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

International Nuclear Information System (INIS)

Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

1990-01-01

Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective of this paper is to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems

Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

International Nuclear Information System (INIS)

Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

1990-03-01

Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective was to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems. 2 refs., 4 figs., 5 tabs

Nuclear Security for Floating Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Skiba, James M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scherer, Carolynn P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-10-13

Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology are proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states

Nuclear power plants in the world - 2010 edition

International Nuclear Information System (INIS)

2010-01-01

This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. These data come from the IAEA's PRIS and AREVA-CEA's GAIA databases. The following aspects are reviewed: 2009 highlights, Main characteristics of reactor types, Map of the French nuclear power plants on 2010/01/01, Worldwide status of nuclear power plants (12/31/2009), Units distributed by countries, Nuclear power plants connected to the Grid- by reactor type groups, Nuclear power plants under construction on 2009, Evolution of nuclear power plants capacities connected to the grid, First electric generations supplied by a nuclear unit in each country, Electrical generation from nuclear power plants by country at the end 2009, Performance indicator of french PWR units, Evolution of the generation indicators worldwide by type, Nuclear operator ranking according to their installed capacity, Units connected to the grid by countries at 12/31/2009, Status of licence renewal applications in USA, Nuclear power plants under construction at 12/31/2009, Shutdown reactors, Exported nuclear capacity in net MWe, Exported and national nuclear capacity connected to the grid, Exported nuclear power plants under construction, Exported and national nuclear capacity under construction, Nuclear power plants ordered at 12/31/2009, Long term shutdown units at 12/31/2009, COL applications in the USA, Recycling of Plutonium in reactors and experiences, Mox licence plants projects, Appendix - historical development, Meaning of the used acronyms, Glossary

Methodology for safety classification of PWR type nuclear power plants items

International Nuclear Information System (INIS)

Oliveira, Patricia Pagetti de

1995-01-01

This paper contains the criteria and methodology which define a classification system of structures, systems and components in safety classes according to their importance to nuclear safety. The use of this classification system will provide a set of basic safety requirements associated with each safety class specified. These requirements, when available and applicable, shall be utilized in the design, fabrication and installation of structures, systems and components of Pressurized Water Reactor Nuclear Power Plants. (author). 13 refs, 1 tab

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E., E-mail: thiago.brito86@yahoo.com.br, E-mail: cabol@ufpe.br, E-mail: wagner@unicap.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Universidade Catolica de Pernambuco (UNICAP), Recife, PE (Brazil). Centro de Ciencias e Tecnologia

2017-11-01

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

International Nuclear Information System (INIS)

Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E.; Universidade Catolica de Pernambuco

2017-01-01

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

Technology development on the assessment of structural integrity of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Seo, Jeong Moon; Choun, Y. S.; Choi, I. K. and others

1999-04-01

Nuclear power plants in Korea show drop off in their performance and safety margin as the age of plants increase. The reevaluation of Kori-1 Unit on its performance and safety for life extension is expected in the near future. However, technologies and information related are insufficient to quantitatively estimate them. The final goal of this study is to develop the basic testing and evaluation techniques related with structural integrity of important nuclear equipment and structures. A part of the study includes development of equipment qualification technique. To ensure the structural integrity of structures, systems, and equipment in nuclear power plants, the following 5 research tasks were performed in the first year. - Analysis of dynamic characteristics of reactor internals - Analysis of engineering characteristics of instrumental earthquakes recorded in Korea - Analysis of ultimate pressure capacity and failure mode of containments building - Development of advanced NDE techniques using ultrasonic resonance scattering - Development of equipment qualification technique against vibration aging. These technologies developed in this study can be used to ensure the structural safety of operational nuclear power plants, and for the long-term life management. (author)

Nuclear plant life - A business decision

International Nuclear Information System (INIS)

Joosten, J.K.

1995-01-01

Regarding the future of the nuclear power option, many scenarios have been put forth over the years. The most commonly accepted projections for installed nuclear capacity show it growing at a rate of about 2% per year throughout the next few decades. These projections appear modes on the surface. However, underlying the projections are critical assumptions and sometimes misconceptions about the lifetimes of existing reactors and how they are determined. The notion of a 40 year plant life is very common. Consequently, many projections start either with the assumption that no plants will be retired in the near terms or with the assumption that each retired plant will be replaced by another nuclear plant after 40 years. Effectively, these assumptions yield future projections for installed capacity that might be characterized as low growth, medium growth and high growth scenarios - or grow, grow, grow. The question remains as to whether or not these assumptions accurately model the driving forces and constraints to nuclear development. After all, there is no scientific basis for believing that all plants, PWRs BWRs, RBMKs, etc., should have the same 40 year life. Most power plant owners purchase the plant for the main reason of supplying electrical power to their consumer. For these owners, electricity production is a day to day commercial activity with various alternatives on how to achieve the prime objective. The decision of which electricity generation alternative to select (gas, coal, nuclear or renewable energy) and how long to operate the plant before replacing it with a new one is essentially a business decision. The paper discusses ageing, the nuclear plant life decision process, the factors which influence the decision and their ramifications regarding the near term growth of nuclear power capacity. The modelling of nuclear plant lifetimes is also discussed. (author). 5 refs, 10 figs, 1 tab

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

International Nuclear Information System (INIS)

Woo, H.H.; Lu, S.C.

1981-01-01

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

Nuclear plant analyzer development and analysis applications

International Nuclear Information System (INIS)

Laats, E.T.

1984-01-01

The Nuclear Plant Analyzer (NPA) is being developed as the U.S. Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. The NPA integrates the NRC's computerized reactor behavior simulation codes such as RELAP5 and TRAC-BWR, both of which are well-developed computer graphics programs, and large repositories of reactor design and experimental data. Utilizing the complex reactor behavior codes as well as the experiment data repositories enables simulation applications of the NPA that are generally not possible with more simplistic, less mechanistic reactor behavior codes. These latter codes are used in training simulators or with other NPA-type software packages and are limited to displaying calculated data only. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

Owners of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Hudson, C.R.; White, V.S.

1996-11-01

Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of July 1996. The report is divided into sections representing different aspects of nuclear plant ownership.

Elecnuc. Nuclear power plants in the world

International Nuclear Information System (INIS)

2000-01-01

This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1999 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1999; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; grid connection forecasts; world electric power market; electronuclear owners and share holders in EU, capacity and load factor; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; 1999 gross load factor by operator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

Advanced plant design recommendations from Cook Nuclear Plant experience

International Nuclear Information System (INIS)

Zimmerman, W.L.

1993-01-01

A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested

Concrete containment vessels (CCV) for nuclear power plants, (1)

International Nuclear Information System (INIS)

Ibe, Yukimi; Kitajima, Masatake

1977-01-01

Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

System and Software Design for the Plant Protection System for Shin-Hanul Nuclear Power Plant Units 1 and 2

International Nuclear Information System (INIS)

Hwang, In Seok; Kim, Young Geul; Choi, Woong Seock; Sohn, Se Do

2015-01-01

The Reactor Protection System(RPS) protects the core fuel design limits and reactor coolant system pressure boundary for Anticipated Operational Occurrences (AOOs), and provides assistance in mitigating the consequences of Postulated Accidents (PAs). The ESFAS sends the initiation signals to Engineered Safety Feature - Component Control System (ESF-CCS) to mitigate consequences of design basis events. The Common Q platform Programmable Logic Controller (PLC) was used for Shin-Wolsung Nuclear Power Plant Units 1 and 2 and Shin-Kori Nuclear Power Plant Units 1, 2, 3 and 4 since Digital Plant Protection System (DPPS) based on Common Q PLC was applied for Ulchin Nuclear Power Plant Units 5 and 6. The PPS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) was developed using POSAFE-Q PLC for the first time for the PPS. The SHN1 and 2 PPS was delivered to the sites after completion of Man Machine Interface System Integrated System Test (MMIS-IST). The SHN1 and 2 PPS was developed to have the redundancy in each channel and to use the benefits of POSAFE-Q PLC, such as diagnostic and data communication. The PPS application software was developed using ISODE to minimize development time and human errors, and to improve software quality, productivity, and reusability

System and Software Design for the Plant Protection System for Shin-Hanul Nuclear Power Plant Units 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Hwang, In Seok; Kim, Young Geul; Choi, Woong Seock; Sohn, Se Do [KEPCO EnC, Daejeon (Korea, Republic of)

2015-10-15

The Reactor Protection System(RPS) protects the core fuel design limits and reactor coolant system pressure boundary for Anticipated Operational Occurrences (AOOs), and provides assistance in mitigating the consequences of Postulated Accidents (PAs). The ESFAS sends the initiation signals to Engineered Safety Feature - Component Control System (ESF-CCS) to mitigate consequences of design basis events. The Common Q platform Programmable Logic Controller (PLC) was used for Shin-Wolsung Nuclear Power Plant Units 1 and 2 and Shin-Kori Nuclear Power Plant Units 1, 2, 3 and 4 since Digital Plant Protection System (DPPS) based on Common Q PLC was applied for Ulchin Nuclear Power Plant Units 5 and 6. The PPS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) was developed using POSAFE-Q PLC for the first time for the PPS. The SHN1 and 2 PPS was delivered to the sites after completion of Man Machine Interface System Integrated System Test (MMIS-IST). The SHN1 and 2 PPS was developed to have the redundancy in each channel and to use the benefits of POSAFE-Q PLC, such as diagnostic and data communication. The PPS application software was developed using ISODE to minimize development time and human errors, and to improve software quality, productivity, and reusability.

Elecnuc. Nuclear power plants worldwide

International Nuclear Information System (INIS)

1998-01-01

This small folder presents a digest of some useful information concerning the nuclear power plants worldwide and the situation of nuclear industry at the end of 1997: power production of nuclear origin, distribution of reactor types, number of installed units, evolution and prediction of reactor orders, connections to the grid and decommissioning, worldwide development of nuclear power, evolution of power production of nuclear origin, the installed power per reactor type, market shares and exports of the main nuclear engineering companies, power plants constructions and orders situation, evolution of reactors performances during the last 10 years, know-how and development of nuclear safety, the remarkable facts of 1997, the future of nuclear power and the energy policy trends. (J.S.)

Operation of Finnish nuclear power plants

International Nuclear Information System (INIS)

Tossavainen, K.

1991-02-01

During the third quarter of 1990 the Finnish nuclear power plant units Loviisa 1 and 2 and TVO I and II were in commercial operation for most of the time. The annual maintenance outages of the Loviisa plant units were held during the report period. All events during this quarter are classified as Level hero (Below Scale) on the International Nuclear Event Scale. Occupational radiation doses and external releases of radioactivity were below authorised limits. Only small amounts of radioactive substances originating in nuclear power plants were detected in samples taken in the vicinity of nuclear power plants

Operation of Finnish nuclear power plants

International Nuclear Information System (INIS)