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Sample records for nuclear plant piping

  1. 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

  2. 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)

  3. Pipe support optimization in nuclear power plants

    International Nuclear Information System (INIS)

    Cleveland, A.B.; Kalyanam, N.

    1984-01-01

    A typical 1000 MWe nuclear power plant consists of 80,000 to 100,000 feet of piping which must be designed to withstand earthquake shock. For the required ground motion, seismic response spectra are developed for safety-related structures. These curves are used in the dynamic analysis of piping systems with pipe-stress analysis computer codes. To satisfy applicable Code requirements, the piping systems also require analysis for weight, thermal and possibly other lasting conditions. Bechtel Power Corporation has developed a design program called SLAM (Support Location Algorithm) for optimizing pipe support locations and types (rigid, spring, snubber, axial, lateral, etc.) while satisfying userspecified parameters such as locations, load combinations, stress and load allowables, pipe displacement and cost. This paper describes SLAM, its features, applications and benefits

  4. Nuclear power plant piping prefabrication and assembly

    International Nuclear Information System (INIS)

    Schmidt, H.

    1990-01-01

    The piping design for nuclear power plants projects reveals, at the beginning, a modification through the application of new fabrication techniques for prefabrication and assembly. This report presents a fabrication methodology which aims to minimize the fabrication and assembly costs as well as to improve and assure quality. (Author) [es

  5. Nuclear power plant piping damping parametric effects

    International Nuclear Information System (INIS)

    Ware, A.G.

    1983-01-01

    The present NRC guidelines for structural damping to be used in the dynamic stress analyses of nuclear power plant piping systems are generally considered to be overly conservative. As a result, plant designers have in many instances used a considerable number of seismic supports to keep stresses calculated by large scale piping computer codes below the allowable limits. In response to this problem, the NRC and EG and G Idaho are engaged in programs to evaluate piping system damping, in order to provide more realistic and less conservative values to be used in seismic analyses. To generate revised guidelines, solidly based on technical data, new experimental data need to be generated and assessed, and the parameters which influence piping system damping need to be quantitatively identified. This paper presents the current state-of-the-art knowledge in the United States on parameters which influence piping system damping. Examples of inconsistencies in the data and areas of uncertainty are explained. A discussion of programs by EG and G Idaho and other organizations to evaluate various effects is included, and both short and long range goals of the program are outlined

  6. Development of support system for nuclear power plant piping

    International Nuclear Information System (INIS)

    Horino, Satoshi

    1987-01-01

    Ishikawajima-Harima Heavy Industries Co., Ltd. has advanced the development of Integrated Nuclear Plant Piping System (INUPPS) for nuclear power plants since 1980, and continued its improvement up to now. This time as its component, a piping support system (PISUP) has been developed. The piping support system deals with the structures such as piping supports and the stands for maintenance and inspection, and as for standard supporting structures, it builds up automatically the structures including the selection of optimum members by utilizing the standard patterns in cooperation with the piping design system including piping stress analysis. As for the supporting structures deviating from the standard, by amending a part of the standard patterns in dialogue from, structures can be built up. By using the data produced in this way, this system draws up consistently a design book, production management data and so on. From the viewpoint of safety, particular consideration is given to the aseismatic capability of nuclear power plants, and piping is fundamentally designed regidly to avoid resonance. It is necessary to make piping supports so as to have sufficient strength and rigidity. The features of the design of piping supports for nuclear power plant, the basic concept of piping support system, the constitution of the software and hardware, the standard patterns and the structural patterns of piping support system and so on are described. (Kako, I.)

  7. Analysis of Defective Pipings in Nuclear Power Plants and Applications of Guided Ultrasonic Wave Techniques

    International Nuclear Information System (INIS)

    Koo, Dae Seo; Cheong, Yong Moo; Jung, Hyun Kyu; Park, Chi Seung; Park, Jae Suck; Choi, H. R.; Jung, S. S.

    2006-07-01

    In order to apply the guided ultrasonic techniques to the pipes in nuclear power plants, the cases of defective pipes of nuclear power plants, were investigated. It was confirmed that geometric factors of pipes, such as location, shape, and allowable space were impertinent for the application of guided ultrasonic techniques to pipes of nuclear power plants. The quality of pipes, supports, signals analysis of weldment/defects, acquisition of accurate defects signals also make difficult to apply the guided ultrasonic techniques to pipes of nuclear power plants. Thus, a piping mock-up representing the pipes in the nuclear power plants were designed and fabricated. The artificial flaws will be fabricated on the piping mock-up. The signals of guided ultrasonic waves from the artificial flaws will be analyzed. The guided ultrasonic techniques will be applied to the inspection of pipes of nuclear power plants according to the basis of signals analysis of artificial flaws in the piping mock-up

  8. Calculation of dynamic hydraulic forces in nuclear plant piping systems

    International Nuclear Information System (INIS)

    Choi, D.K.

    1982-01-01

    A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)

  9. Piping support load data base for nuclear plants

    International Nuclear Information System (INIS)

    Childress, G.G.

    1991-01-01

    Nuclear Station Modifications are continuous through the life of a Nuclear Power Plant. The NSM often impacts an existing piping system and its supports. Prior to implementation of the NSM, the modified piping system is qualified and the qualification documented. This manual review process is tedious and an obvious bottleneck to engineering productivity. Collectively, over 100,000 piping supports exist at Duke Power Company's Nuclear Stations. Engineering support must maintain proper documentation of all data for each support. Duke Power Company has designed and developed a mainframe based system that: directly uses Support Load Summary data generated by a piping analysis computer program; streamlines the pipe support evaluation process; easily retrieves As-Built and NSM information for any pipe support from an NSM or AS-BUILT data base; and generated documentation for easy traceability of data to the information source. This paper discusses the design considerations for development of Support Loads Database System (SLDB) and reviews the program functionality through the user menus

  10. Manufacture of piping components for nuclear power plants

    International Nuclear Information System (INIS)

    Bartecek, R.

    1983-01-01

    Hammer forging of hollow forging ingots, extrusion and elestroslag remelting may be used for the manufacture of large pipes. Technologies have been developed for the manufacture of elbows based on various types of forming. These procedures mainly include the hydraulic pressing of elbows from tubes and the pressing of symmetrical halves of elbows with subsequent welding. The hammer forging of valves, cross pieces, etc., has been replaced by forging and pressing. In order to prevent failures from occurring in the pipes during operation of nuclear power plants, pipes are being made of larger forgings, which reduces the number of welds. This improves the quality of the pipes, reduces production and assembly costs and is metal-saving. (E.S.)

  11. Nuclear power plant piping damping parametric effects

    International Nuclear Information System (INIS)

    Ware, A.G.

    1983-01-01

    The NRC and EG and G Idaho are engaged in programs to evaluate piping-system damping, in order to provide realistic and less conservative values to be used in seismic analyses. To generate revised guidelines, solidly based on technical data, new experimental data need to be generated and assessed, and the parameters which influence piping-system damping need to be quantitatively identified. This paper presents the current state-of-the-art knowledge in the United States on parameters which influence piping-system damping. Examples of inconsistencies in the data and areas of uncertainty are explained. A discussion of programs by EG and G Idaho and other organizations to evaluate various effects are included, and both short-and long-range goals of the program are outlined

  12. Pipe line systems in nuclear power plant

    International Nuclear Information System (INIS)

    Sasada, Yasuhiro; Tanno, Kazuo; Shibato, Eizo.

    1979-01-01

    Purpose: To prevent stress corrosion cracks, in particular, for branched pipeways by conducting water quality control in the branched pipeways as well as in the main pipeways, and reducing the thermal stress in the branched pipeways. Constitution: A water quality monitoring device is provided to a drain pipe and a failed element detection pipe to monitor the quality of stagnated water continuously or periodically. If the impurity concentration or oxygen concentration exceeds a specified value in the stagnated water, a drain valve or a check valve is opened by a signal from the water quality monitoring device to replace the stagnated water with recycling water in the main pipeway. The temperature for the branched loop pipeway and the main pipeway are collectively kept to a same temperature to thereby reduce the thermal stress in the branched pipeway. (Kawakami, Y.)

  13. Development of automatic pipe welder for nuclear power plant

    International Nuclear Information System (INIS)

    Iwamoto, Taro; Ando, Shimon; Omae, Tsutomu; Ito, Yoshitoshi; Araya, Takeshi.

    1978-01-01

    Numerous pipings are installed in nuclear power plants, and of course, the reliability of these pipings are very important to preserve the safety of the plants. These pipings undergo periodic inspection yearly, and when some defects are found or some reconstructions to superior systems are made, field welding in the plants is required. When the places to be welded are in containment vessels, the works must be carried out in radiation environment. In order to maintain the highest quality of welding and to reduce the radiation exposure of workers, many skilled workers are required. This automatic pipe welder was developed to solve these problems, aiming at carrying out welding works by remote control at the safe places outside containment vessels. Especially in order to obtain the highest quality of welding, it was not perfectly automated, but the man-machine system so as to enable to utilize the delicate sense of workers was adopted. The visual and contact detecting systems to monitor welding works, remote control system, computer control, light, small and easily installed welding head, grinding and supersonic flow detecting equipments, the power source of transistor switching type, air cooling equipment, and the function for setting welding conditions according to algorithm were added to the welding machine. The outline and main components of this automatic pipe welder are explained. (Kako, I.)

  14. Piping information centralized management system for nuclear plant, PIMAS

    International Nuclear Information System (INIS)

    Matsumoto, Masaru

    1977-01-01

    Piping works frequently cause many troubles in the progress of construction works, because piping is the final procedure in design and construction and is forced to suffer the problems in earlier stages. The enormous amount of data on quality control and management leads to the employment of many unskilled designers of low technical ability, and it causes confusion in installation and inspection works. In order to improve the situation, the ''piping information management system for nuclear plants (PIMAS)'' has been introduced attempting labor-saving and speed-up. Its main purposes are the mechanization of drafting works, the centralization of piping informations, labor-saving and speed-up in preparing production control data and material management. The features of the system are as follows: anyone can use the same informations whenever he requires them because the informations handled in design works are contained in a large computer; the system can be operated on-line, and the terminals are provided in the sections which require informations; and the sub-systems are completed for preparing a variety of drawings and data. Through the system, material control has become possible by using the material data in each plant, stock material data and the information on the revision of drawings in the design department. Efficiency improvement and information centralization in the manufacturing department have also been achieved because the computer has prepared many kinds of slips based on unified drawings and accurate informations. (Wakatsuki, Y.)

  15. Fatigue evaluation of socket welded piping in nuclear power plant

    International Nuclear Information System (INIS)

    Vecchio, R.S.

    1996-01-01

    Fatigue failures in piping systems occur, almost without exception, at the welded connections. In nuclear power plant systems, such failures occur predominantly at the socket welds of small diameter piping ad fillet attachment welds under high-cycle vibratory conditions. Nearly all socket weld fatigue failures are identified by leaks which, though not high in volume, generally are costly due to attendant radiological contamination. Such fatigue cracking was recently identified in the 3/4 in. diameter recirculation and relief piping socket welds from the reactor coolant system (RCS) charging pumps at a nuclear power plant. Consequently, a fatigue evaluation was performed to determine the cause of cracking and provide an acceptable repair. Socket weld fatigue life was evaluated using S-N type fatigue life curves for welded structures developed by AASHTO and the assessment of an effective cyclic stress range adjacent to each socket weld. Based on the calculated effective tress ranges and assignment of the socket weld details to the appropriate AASHTO S-N curves, the socket weld fatigue lives were calculated and found to be in excellent agreement with the accumulated cyclic life to-date

  16. Damping values for nuclear power plant piping during seismic events and fluid-induced transients

    International Nuclear Information System (INIS)

    Ware, A.G.

    1986-01-01

    For several years the Idaho National Engineering Laboratory (INEL) has been assisting the United States Nuclear Regulatory Commission (USNRC) in efforts to establish best-estimate damping values for use in the dynamic analysis of nuclear power plant piping systems. Data from a number of piping vibration tests conducted at facilities worldwide (including the INEL) have been collected, evaluated, reported, and placed in a nuclear piping data bank at the INEL. These data are being used to justify changes in allowable damping values for use in nuclear piping design, thus making piping systems safer, less costly, and easier to inspect and maintain

  17. An assessment of seismic margins in nuclear plant piping

    International Nuclear Information System (INIS)

    Chen, W.P.; Jaquay, K.R.; Chokshi, N.C.; Terao, D.

    1995-01-01

    Interim results of an ongoing program to assist the U.S. Nuclear Regulatory Commission (NRC) in developing regulatory positions on the seismic analyses of piping and overall safety margins of piping systems are reported. Results of reviews of previous seismic testing, primarily the Electric Power Research Institute (EPRI)/NRC Piping and Fitting Dynamic Reliability Program, and assessments of the ASME Code, Section III, piping seismic design criteria as revised by the 1994 Addenda are reported. Major issues are identified herein only. Technical details are to be provided elsewhere. (author). 4 refs., 2 figs

  18. Dimensional control of buttwelding pipe fitting for nuclear power plant Class 1 piping systems

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Moore, S.E.; Robinson, J.N.

    1976-11-01

    Dimensional controls of wrought steel buttwelding fittings are examined from the standpoint of design adequacy. A fairly large number of fittings were purchased from different manufacturers. The dimensions of each fitting were measured and correlated along with additional information obtained from the manufacturers in an effort to establish ''standard'' shapes. This information and a critical examination of the present ANSI standards is used to develop a ''Supplementary Standard.'' The Supplementary Standard is intended to provide improved dimensional control and more complete design information for fittings used in Class 1 nuclear power plant piping systems

  19. Mechanized ultrasonic examination of piping systems in nuclear power plants

    International Nuclear Information System (INIS)

    Edelmann, X.; Pfister, O.; Allidi, F.

    1988-01-01

    The success of mechanized ultrasonic examination applied on welds in piping systems in nuclear power plants is highly dependent on its careful preparation. From the development of an adequate examination technique to its implementation on site, many problems are to be solved. This is especially the case when dealing with austenitic welds or dissimilar metal welds. In addition to the specific needs for examination technique based on material properties and requirements for minimum flaw size detection, accessibility and radiation aspects have to be considered. A crew of skilled and highly trained examination personnel is required. Experience in various nuclear power plants, - BWR's and PWR's of different designs - has shown, that even difficult examination problems can be successfully solved, provided that there is a good preparation. The necessary step by step proceeding is illustrated by examples concerning mechanized examination. Preservice inspections and in-service inspections with specific requirements, due to the types of flaws to be found or the type of material concerned, are discussed

  20. Noncondensable gas accumulation phenomena in nuclear power plant piping

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Aoki, Kazuyoshi; Sato, Teruaki; Shida, Akira; Ichikawa, Nagayoshi; Nishikawa, Akira; Inagaki, Tetsuhiko

    2011-01-01

    In the case of the boiling water reactor, hydrogen and oxygen slightly exist in the main steam, because these noncondensable gases are generated by the radiolytic decomposition of the reactor water. BWR plants have taken measures to prevent noncondensable gas accumulation. However, in 2001, the detonation of noncondensable gases occurred at Hamaoka-1 and Brunsbuttel, resulting in ruptured piping. The accumulation phenomena of noncondensable gases in BWR closed piping must be investigated and understood in order to prevent similar events from occurring in the future. Therefore, an experimental study on noncondensable gas accumulation was carried out. The piping geometries for testing were classified and modeled after the piping of actual BWR plants. The test results showed that 1) noncondensable gases accumulate in vertical piping, 2) it is hard for noncondensable gases to accumulate in horizontal piping, and 3) noncondensable gases accumulate under low-pressure conditions. A simple accumulation analysis method was proposed. To evaluate noncondensable gas accumulation phenomena, the three component gases were treated as a mixture. It was assumed that the condensation amount of the vapor is small, because the piping is certainly wrapped with heat insulation material. Moreover, local thermal equilibrium was assumed. This analysis method was verified using the noncondensable gas accumulation test data on branch piping with a closed top. Moreover, an experimental study on drain trap piping was carried out. The test results showed that the noncondensable gases dissolved in the drain water were discharged from the drain trap, and Henry's law could be applied to evaluate the amount of dissolved noncondensable gases in the drain water. (author)

  1. Hydraulic simulation of the systems of a nuclear power plant for charges calculation in piping

    International Nuclear Information System (INIS)

    Masriera, N.

    1990-01-01

    This work presents a general description of the methodology used by the ENACE S.A. Fluids Working Group for hydraulics simulation of a nuclear power plant system for the calculation charges in piping. (Author) [es

  2. Nuclear power plant pressure vessels. Control of piping

    International Nuclear Information System (INIS)

    2000-01-01

    The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

  3. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    International Nuclear Information System (INIS)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik; Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae

    2015-01-01

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  4. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik [School of Materials Science and Engineering, Andong National University, Andong (Korea, Republic of); Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae [Power Engineering Research Institute, KEPCO Engineering and Construction Company, Seongnam (Korea, Republic of)

    2015-02-15

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  5. Study on quality control measures of static casting main pipe in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zhenbiao; Li Guanying; Liu Zhicheng

    2013-01-01

    This study analyzes the main reasons which impact the quality of primary pipe static casting elbows in PWR-M310 nuclear power plant. The quality control measures are developed from the election and inspection of material, improving sand production and casting process, improving lean management of personnel. The static casting defects of primary pipe elbows for Fuqing Unit 1 and 2 were down to less than 50% of the former project. The quality of static casting for the primary pipe elbows was significantly improved. Moreover, the implementation saves human resources and financing to repair casting defects, and also helps to win the delivery schedule. The quality control measures are good reference for improving primary pipe casting process. This study provides valuable experience for further study of improving the quality of static casting for the primary pipe of PWR nuclear power plant. (authors)

  6. Review and assessment of research relevant to design aspects of nuclear power plant piping systems. Final report

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Maxey, W.A.; Eiber, R.J.

    1977-06-01

    Significant research on piping systems is evaluated, and the correlation of that research with design practices is presented. The objective is to quantify the research/design practices in terms of the reliability of piping used in nuclear power plants

  7. Technical considerations for flexible piping design in nuclear power plants

    International Nuclear Information System (INIS)

    Lu, S.C.; Chou, C.K.

    1985-01-01

    The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. A couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design were investigated. It was concluded that these changes substantially reduce calculated piping responses and allows piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements

  8. Nuclear Power Plants Secondary Circuit Piping Wall-Thinning Management in China

    International Nuclear Information System (INIS)

    Zhong Zhimin; Li Jinsong; Zheng Hui

    2012-01-01

    Research and field feedbacks showed that nuclear power plants secondary circuit steam and water piping are more sensitive than that of fuel plant to the attack of flow-accelerated corrosion (FAC). FAC, Liquid droplet impingement or cavitation erosion will cause secondary circuit piping local wall-thinning in NPPs. Without effective management, the wall-thinning in those high energy piping will cause leakage or pipe rupture during nuclear power plant operation, more seriously cause unplanned shut down, injured and fatality, or heavy economic losses. This paper briefly introduces the history, development and state of the art of secondary circuit piping wall-thinning management in China NPPs. Then, the effectiveness of inspection grid size selecting was analyzed in detail based on field feedbacks. EPRI recommendatory inspection grid, JSME code recommendatory grid and plant specific inspection grid were compared and the detection probabilities of local wall-thinning were estimated. Then, the development and application of NPPs Secondary Circuit Piping Wall Thickness Management Information System, developed, operated and maintained by our team, was briefly introduced and the statistical analysis results of 11 PWR units were shared. It was conclude that the long term, systemic, effective wall-thinning management strategy of high energy piping was very important to the safety and economic operation of NPPs. Furthermore, take into account the actual situation of China nuclear power plants, some advice and suggestion on developing effective nuclear power plant secondary circuit steam and water piping wall-thinning management system are put forward from code development, design and manufacture, operation management, pipeline and locations selection, inspection method selection and application, thickness measurement result evaluation, residual life predication and decision making, feedbacks usage, personnel training and etc. (author)

  9. Seismic design of equipment and piping systems for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi

    1997-01-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  10. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  12. Development of Structural Health Monitoring System for pipes in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Eom, H. S.; Choi, Y. C.; Shin, S. H.; Youn, D. B.; Park, J. H.

    2010-01-01

    Structural health monitoring (SHM) has becoming an important issue in the maintenance of various structures such as large steel plates, vessels, and pipes in nuclear power plants. There are important factors to be considered in developing an SHM system. With consideration of these factors, we have developed a computerized multi-channel ultrasonic system that can handle array transducers and generate a high-power pulse for online SHM of the plates and pipes. The proposed system is compact but has all the necessary functions for SHM of important structure such as pipes and plates in a NPP

  13. 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.

  14. Inspection of nuclear power plant piping welds by in-process acoustic emission monitoring

    International Nuclear Information System (INIS)

    Prine, D.W.

    1976-01-01

    The results of using in-process acoustic emission monitoring on nuclear power plant piping welds are discussed. The technique was applied to good and intentionally flawed test welds as well as production welds, and the acoustic emission results are compared to standard NDT methods and selected metallographic cross-sections

  15. An overview of environmental degradation of materials in nuclear power plant piping systems

    International Nuclear Information System (INIS)

    Shack, W.J.

    1988-01-01

    Several types of environmental degradation of piping in light water reactor (LWR) power systems have already had significant economic impact on the industry. These include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping, erosion-corrosion of carbon steel piping in secondary systems, and a variety of types of fatigue failures. In addition, other problems have been identified that must be addressed in considering extended lifetimes for nuclear plants. These include the embrittlement of cast stainless steels after extended thermal aging at reactor operating temperatures and the effect of reactor environments on the design margin inherent in the ASME Section III fatigue design curves especially for carbon steel piping. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

  16. Qualification of Manual Phased Array Ultrasonic Techniques for Pipe Weld Inspection in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Poirier, J.; Hayes, P.; Vicat, F. [GE Inspection Technologies (United States)

    2011-07-01

    Phasor XS can be used for piping weld inspection in any facilities that use EPRI procedures (example: nuclear power plant in Usa, Japan, ...). Whole pipe range is inspected with 5 probes and 6 wedges: 4 1-dimensional probe for sound wave scanning (different frequency, different apertures); 1 dual matrix probe for LW scanning; there are 3 types of wedges optimized for weld inspection. Weld is scanned in 'Raster Scan', maximum range from 35 up to 80 degrees. Probe selection is defined in the procedure according to pipe diameter, pipe thickness and type of access (single or dual side). We have to note that datasets for dual matrix probe are provided with the procedure because this kind of probe cannot be programmed inside Phasor XS

  17. Evaluation and summary of seismic response of above ground nuclear power plant piping to strong motion earthquakes

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1985-01-01

    The purpose of this paper is to summarize the observations and experience which has been developed relative to the seismic behavior of above-ground, building-supported, industrial type piping (similar to piping used in nuclear power plants) in strong motion earthquakes. The paper also contains observations regarding the response of piping in experimental tests which attempted to excite the piping to failure. Appropriate conclusions regarding the behavior of such piping in large earthquakes and recommendations as to future design of such piping to resist earthquake motion damage are presented based on observed behavior in large earthquakes and simulated shake table testing

  18. Estimation of leak rate through circumferential cracks in pipes in nuclear power plants

    Directory of Open Access Journals (Sweden)

    Jai Hak Park

    2015-04-01

    Full Text Available The leak before break (LBB concept is widely used in designing pipe lines in nuclear power plants. According to the concept, the amount of leaking liquid from a pipe should be more than the minimum detectable leak rate of a leak detection system before catastrophic failure occurs. Therefore, accurate estimation of the leak rate is important to evaluate the validity of the LBB concept in pipe line design. In this paper, a program was developed to estimate the leak rate through circumferential cracks in pipes in nuclear power plants using the Henry–Fauske flow model and modified Henry–Fauske flow model. By using the developed program, the leak rate was calculated for a circumferential crack in a sample pipe, and the effect of the flow model on the leak rate was examined. Treating the crack morphology parameters as random variables, the statistical behavior of the leak rate was also examined. As a result, it was found that the crack morphology parameters have a strong effect on the leak rate and the statistical behavior of the leak rate can be simulated using normally distributed crack morphology parameters.

  19. In-service diagnostics of main circulating circuit pipes of WWER nuclear power plants

    International Nuclear Information System (INIS)

    Svoboda, V.; Merta, J.; Merta, V.

    1982-01-01

    The application is discussed of the acoustic emission method for testing the integrity of the components of the main circulating circuit of the WWER 440 nuclear power plant. A description is given of the main circulating circuit and a stress analysis on the basis of strength computations considering operating modes is presented. An analysis is also presented of the possible damage of the pipe material as related to the application of the acoustic emission method for in-service inspection of the pipes. Certain practical problems of application are discussed. (author)

  20. 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

  1. Integrated CAE system for nuclear power plants. Development of piping design check system

    International Nuclear Information System (INIS)

    Narikawa, Noboru; Sato, Teruaki

    1994-01-01

    Toshiba Corporation has developed and operated the integrated CAE system for nuclear power plants, the core of which is the engineering data base to manage accurately and efficiently enormous amount of data on machinery, equipment and piping. As the first step of putting knowledge base system to practical use, piping design check system has been developed. By automatically checking up piping design, this system aims at the prevention of overlooking mistakes, efficient design works and the overall quality improvement of design. This system is based on the thought that it supports designers, and final decision is made by designers. This system is composed of the integrated data base, a two-dimensional CAD system and three-dimensional CAD system. The piping design check system is one of the application systems of the integrated CAE system. Object-oriented programming is the base of the piping design check system, and design knowledge and CAD data are necessary. As to the method of realizing the check system, the flow of piping design, the checkup functions, the checkup of interference and attribute base, and the integration of the system are explained. (K.I)

  2. 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

  3. Estimation of Leak Rate Through Cracks in Bimaterial Pipes in Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Jai Hak Park

    2016-10-01

    Full Text Available The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry–Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.

  4. Nuclear power plant steam pipes repairing with TIRANT 3 robot system

    International Nuclear Information System (INIS)

    Soto Tomas, Marcelo; Curiel Nieva, Marceliano; Monzo Blasco, Enrique; Rodriguez, Salvador Pineda; Vaquer Perez, Juan I.

    2011-01-01

    A typical application functions covering the steam pipes inner surface in coal-fired power station and nuclear power plants. The results of this process are spectacular in terms of protection against corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the postural position (usually kneeling) in small diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and quality assurance. For all these reasons, Grupo Dominguis has developed the TIRANT 3 robot, a worldwide innovative system, for metallization of steam pipes inner surface. TIRANT 3 robot is teleoperated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, TIRANT 3 system permits to increase resulting coating uniformity, and thus its resistance, keeping selected parameters constant depending on required type and thickness of wire. TIRANT 3 system has successfully worked in 2010 during the stops refueling of the Units I and II of Laguna Verde nuclear power plant in Mexico. (author)

  5. Numerical and experimental analysis of the vibratory behavior of a nuclear power plant piping system excitated by a pump

    International Nuclear Information System (INIS)

    Vatin, E.; Guillou, J.; Tephany, F.; Trollat, C.

    1993-08-01

    This paper presents a study on the dynamic response of piping systems installed in the French 1300 MWe Nuclear Power Plants. Variations in pressure are generated by a multi-staged centrifugal pump mounted on the piping system and provide a dynamic excitation of the pipe. This type of dynamic loading has led to nozzle cracks for some of the pipes, whereas, for other installations, it has not be found detrimental. This study presents an explanation of the different dynamic behavior observed at the various plants. To this end, a numerical model, calibrated with on-site measurements, is impleted. (authors). 8 figs., 1 tab., 5 refs

  6. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  7. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  8. Predicting local distributions of erosion-corrosion wear sites for the piping in the nuclear power plant using CFD models

    International Nuclear Information System (INIS)

    Ferng, Y.M.

    2008-01-01

    The erosion-corrosion (E/C) wear is an essential degradation mechanism for the piping in the nuclear power plant, which results in the oxide mass loss from the inside of piping, the wall thinning, and even the pipe break. The pipe break induced by the E/C wear may cause costly plant repairs and personal injures. The measurement of pipe wall thickness is a useful tool for the power plant to prevent this incident. In this paper, CFD models are proposed to predict the local distributions of E/C wear sites, which include both the two-phase hydrodynamic model and the E/C models. The impacts of centrifugal and gravitational forces on the liquid droplet behaviors within the piping can be reasonably captured by the two-phase model. Coupled with these calculated flow characteristics, the E/C models can predicted the wear site distributions that show satisfactory agreement with the plant measurements. Therefore, the models proposed herein can assist in the pipe wall monitoring program for the nuclear power plant by way of concentrating the measuring point on the possible sites of severe E/C wear for the piping and reducing the measurement labor works

  9. A review of nondestructive examination technology for polyethylene pipe in nuclear power plant

    Science.gov (United States)

    Zheng, Jinyang; Zhang, Yue; Hou, Dongsheng; Qin, Yinkang; Guo, Weican; Zhang, Chuck; Shi, Jianfeng

    2018-05-01

    Polyethylene (PE) pipe, particularly high-density polyethylene (HDPE) pipe, has been successfully utilized to transport cooling water for both non-safety- and safety-related applications in nuclear power plant (NPP). Though ASME Code Case N755, which is the first code case related to NPP HDPE pipe, requires a thorough nondestructive examination (NDE) of HDPE joints. However, no executable regulations presently exist because of the lack of a feasible NDE technique for HDPE pipe in NPP. This work presents a review of current developments in NDE technology for both HDPE pipe in NPP with a diameter of less than 400 mm and that of a larger size. For the former category, phased array ultrasonic technique is proven effective for inspecting typical defects in HDPE pipe, and is thus used in Chinese national standards GB/T 29460 and GB/T 29461. A defect-recognition technique is developed based on pattern recognition, and a safety assessment principle is summarized from the database of destructive testing. On the other hand, recent research and practical studies reveal that in current ultrasonic-inspection technology, the absence of effective ultrasonic inspection for large size was lack of consideration of the viscoelasticity effect of PE on acoustic wave propagation in current ultrasonic inspection technology. Furthermore, main technical problems were analyzed in the paper to achieve an effective ultrasonic test method in accordance to the safety and efficiency requirements of related regulations and standards. Finally, the development trend and challenges of NDE test technology for HDPE in NPP are discussed.

  10. Structural Health Monitoring of Piping in Nuclear Power Plants - A Review of Efficiency of Existing Methods

    International Nuclear Information System (INIS)

    Stepinski, Tadeusz

    2011-05-01

    In the first part of the report, we review various efforts that have been recently performed in the USA in the field of reactor health monitoring. They were carried out by different organizations and they addressed different issues related to the safety of nuclear reactors. Among other aspects, we present technical issues related to the design of a self-diagnostic monitoring system for the next generation of nuclear reactors. We also give a brief review of the international experience of such systems in today's reactors. In the second part of the report we focus on long range ultrasonic techniques that can be used for monitoring piping in nuclear reactors. Common strategy used in the Swedish nuclear plants is leak before break (LBB), which relies on monitoring leaks from the pipelines as indications of possible pipe break. Significant parts of piping systems are partly or entirely inaccessible for the NDE inspectors, which complicates the use of proactive strategies. One solution to the problem could be implementing monitoring systems capable of monitoring pipelines over a long range. The method, which has shown much promise in such applications is the UT based on guided waves (GW) referred to as long range ultrasound testing (LRUT). In the report we give a brief review of the GW theory followed by the presentation the commercial GW instruments and transducers designed for the LRUT of piping. We also present examples of the baseline based systems using permanently installed transducers. In the final part we report capacity tests of the LRUT instruments performed in collaboration with two different manufactures

  11. Situation of secondary system piping wearing in overseas nuclear power plants

    International Nuclear Information System (INIS)

    Chiba, Goro

    2005-01-01

    In consideration of secondary system piping rupture accident at Mihama Nuclear Power Station Unit 3 of Kansai Electric Power Company in August 2004, the management system of secondary pipe wall thickness of Japan and foreign countries were investigated. Moreover, the tendency of the secondary piping thinning events on overseas which the Institute of Nuclear Safety System, Inc. (INSS) obtained was analyzed in order to verify the validity of the Japanese management system. Consequently, it was shown that in the U.S., the fault phenomenon of secondary system piping was reported continuously, and there were also many cases of both degradation and penetration of pipe wall. (author)

  12. Reliability Data Handbook for Piping Components in Nordic Nuclear Power Plants - R Book, Phase 2

    International Nuclear Information System (INIS)

    Hedtjaern Swaling, Vidar; Olsson, Anders

    2011-02-01

    This report presents results of a long research and development project financed by the regulatory body Straalsaekerhetsmyndigheten (SSM) (former SKI), the Swedish nuclear power plant licensees. The report presents a harmonized method for estimating Reliability Data for Piping Components in ASME code class 1 and 2 piping components (R-Book). Data in the R-Book is measured based on 'data driven' strategy. This first version of the R-Book comprises rupture frequencies and failure rates for all systems where ASME Code Class 1 or 2 events could be found in the OECD OPDE database. Nordic and Non-Nordic data are presented separately. Worldwide experience data is used to set up the relevant calculation cases, i.e. intersections of attributes for which there are at least one event present

  13. Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

    International Nuclear Information System (INIS)

    Charalambus, B.

    1992-01-01

    Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative. (orig.)

  14. Improvement of layout and piping design for PWR nuclear power plants

    International Nuclear Information System (INIS)

    Nozue, Kosei; Waki, Masato; Kashima, Hiroo; Yoshioka, Tsuyoshi; Obara, Ichiro.

    1983-01-01

    For a nuclear power plant, a period of nearly ten years is required from the initial planning stage to commencement of transmission after passing through the design, manufacturing, installation and trial running stages. In the current climate there is a trend that the time required for nuclear power plant construction will further increase when locational problems, thorough explanation to residents in the neighborhood of the construction site and their under-standing, subsequent safety checks and measures to be taken in compliance with various controls and regulations which get tighter year after year, are taken into account. Under such circumstances, in order to satisfy requirements such as improving the reliability of the nuclear power plant design, manufacturing and construction departments, improvements in the economy as well as the quality and shortening of construction periods, the design structure for Mitsubishi PWR nuclear power plants was thoroughly consolidated with regard to layout and piping design. At the same time, diversified design improvements were made with the excellent domestic technology based on plant designs imported from the U.S.A. An outline of the priority items is introduced in this paper. (author)

  15. The water treatment in the dual-purpose nuclear plants of Babcock and Wilcox with straight pipes

    International Nuclear Information System (INIS)

    Martynova, O.I.

    1978-01-01

    A report is given on water processing and water chemistry in the dual-purpose nuclear power plants (as compared to the single-purpose nuclear power plants) of Babcock and Wilcox, with flow steam generators with straight pipes. The most important materials, especially regarding their corrosion resistance, and the water composition during 'hot' start-up of the Okonie-I power plant, the quality factors of the feedwater, the water quality factors of the steam generator with fast start-up and the experience with numerous corrosion-caused defects in steam generator pipes are dealt with from the aspect of optimum water processing and successful continuous operation. (HK) [de

  16. The Analysis of the Field Application Methodology of Electromagnetic Ultrasonic Testing for Piping in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chi Seung; Joo, Keum Jong; Choi, Jung Kweun; Um, Byung Kook; Park, Jea Suk [Korea Advanced Ispection Technology Co., Daejeon (Korea, Republic of)

    2008-08-15

    Nuclear plant piping is classified as the safety class and non-safety class piping in usual. Safety class piping has been examined in accordance with ASME Section XI and V during PSI/ISI using RT, UT, PT, ECT, etc and evaluated periodically for integrity. But failures in piping had reported at non-welded parts and non-safety class pipings as well as the safety class pipings. The existing NDT methods are suitable for the specific parts for instance weldments to inspect but difficult to examine all parts (total coverage) of pipe line and very expensive in cost and consume the time. And also inspection using those methods is difficult and limited for the parts which are complex configuration, embedded under ground and installed at high radiation area in nuclear power plants. In order to inspect all parts of long range piping systems and reduce the inspection time and cost, the electromagnetic ultrasonic inspection technology is suitable and effective. The electromagnetic ultrasonic method can cover more than 50 m apart from sensor at one time without moving the sensor and examined the parts which are in difficulties for accessibility, for example, high radiation area, insulated components and embedded under ground.

  17. Ageing of reinforced concrete pipes subjected to seawater in nuclear plants: optimization of maintenance operations

    International Nuclear Information System (INIS)

    Auge, L.; Capra, B.; Lasne, M.; Benefice, P.; Comby, R.

    2007-01-01

    Seaside nuclear power plants have to face the ageing of nuclear reactor cooling piping systems. In order to minimize the duration of the production unit shutdown, maintenance operations have to be planned well in advance. In a context where owners of infrastructures tend to extend the life span of their goods while having to keep the safety level maximum, it is more and more important to develop high level expertise and know-how in management of infrastructures life cycle. A patented monitoring technique based on optic fiber sensors, has been designed. This preventive maintenance enables the owner to determine criteria for network replacement based on degradation impacts. A methodology to evaluate and optimize operation budgets, depending on predictions of future functional deterioration and available maintenance solutions, has been developed and applied. (authors)

  18. Stress corrosion evaluation on stainless steel 304 pipes in Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    1996-01-01

    Inside the frame of the project IAEA/MEX-41044 'Stress corrosion as a starting event of accidents in nuclear plants', and of the institutional project IA-252 under the same name, it was required from the Laguna Verde Nuclear Plant, material equivalent to the one employed in the piping of the primary recycling system. Laguna Verde Nuclear Plant granted two tracks of tubes, that could be used to substitute the ones that are in operation, as is the tube SA-358TP304 CL-QC with transversal welding, designated as ER-316-LQA. According to the report entitles 'Revision of the operational experience related to corrosion in the nuclear plants' it was found that the stress corrosion is the principal mechanism of corrosion present in the nuclear plants. Previous records indicate that sensitized stainless steels are resistant to stress corrosion in testings of constant loading in sea water (3.5% of chlorides approximately) to 80 Centigrade and to 80% of the limit of conveyance and that a solution of 22% of NaCl to 90 Centigrade, produces cracking due to stress corrosion in highly sensitized steels, in tests of speed of slow extension (SSRT), to a speed of 1x10 -6 s -1 . Daniels reports that there is a direct relation between the speed limit of detection of the SSRT test and the concentration of chlorides, for stainless steels tested to 100 Centigrade. The minimum detection speed of susceptibility to stress corrosion for solution to 20% of NaCl, is of 1x10 -7 s -1 . Taking into account these considerations, the employment of a solution with 22% of NaCl to 90 Centigrade to a speed of 1x10 -6 s -1 seems a good choice for the evaluation of stainless steel. (Author)

  19. Nuclear Power Plant Steam Pipes repairing with Tirant 3R Robot System

    International Nuclear Information System (INIS)

    Ruiz-Martinez, Jose-Tomas; Soto-Tomas, Marcelo; Curiel-Nieva, Marceliano; Monzo-Blasco, Enrique; Pineda-Rodriguez, Salvador; Vaquer-Perez, Juan-Ignacio

    2012-09-01

    The metallization arc spray process is based on the projection of molten metal, supplied by means of different stainless alloys wire, over a surface of carbon steel usually, with the object of serving as protection against flow assisted corrosion (FAC), increasing resistance to abrasion and deteriorations. A typical application functions covering the steam pipes inner surface in Coal-fired power station and Nuclear Power Plants. The results of this process are spectacular in terms of protection against flow assisted corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the worker's postural position (usually kneeling) in 32' diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and Quality Assurance. An increase in the uniformity of the projected coating, increase the resistance and give a better surface protection. For all these reasons, Lainsa has developed the TIRANT 3 R system, a worldwide innovative system, for metallization of steam pipes inner surface. TIRANT 3 R system is tele-operated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, TIRANT 3 R system permits to increase resulting coating uniformity and thus its resistance, keeping selected parameters constant (forward speed, rotation speed and inner surface distance) depending on required type and

  20. Nuclear power plant steam pipes repairing with Tirant 3 Robot system

    Energy Technology Data Exchange (ETDEWEB)

    Soto, M.; Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Lazaro, F. [Revestimientos Anticorrosivos Industriales, S. L. U., Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Arnaldos, A., E-mail: m.soto@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    The metallization arc spray process is based on the projection of molten metal, supplied by means of different stainless alloys wire, over a surface of carbon steel usually, with the object of serving as protection against erosion-corrosion, increasing resistance to abrasion and detrition. A typical application functions covering the steam pipes inner surface in coal-fired power station and nuclear power plants. The results of this process are spectacular in terms of protection against corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the worker's postural position (usually kneeling) in 32 diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting, ... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and quality assurance. An increase in the uniformity of the projected coating, increase the resistance and give a better surface protection. For all these reasons, Lainsa has developed the Tirant 3 robot, a worldwide innovative system, for metallization of steam pipes inner surface. Tirant 3 robot is tele operated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, Tirant 3 system permits to increase resulting coating uniformity and thus its resistance, keeping selected parameters constant (forward speed, rotation speed and inner surface distance) depending on required type and thickness of wire. (Author)

  1. Fatigue damage evaluation of stainless steel pipes in nuclear power plants using positron annihilation lineshape analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kawaguchi, Yasuhiro [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Nakamura, Noriko; Yusa, Satoru [Ishikawajima-Harima Heavy Industries Co., Tokyo (Japan)

    2002-09-01

    Since positron annihilation lineshape analysis can evaluate the degree of fatigue damage by detecting defects such as dislocations in metals, we applied this method to evaluate that in a type 316 stainless steel pipe which was used in the primary system of a nuclear power plant. Using {sup 68}Ge as a positron source, an energy spread of annihilation gamma ray peak from the material was measured and expressed as the S-parameter. Actual plant material cut from a surge line pipe of a pressurizer in a pressurized water reactor type nuclear power plant was measured by positron annihilation lineshape analysis and the S-parameter was obtained. Comparing the S-parameter with a relationship between the S-parameter and fatigue life ratio of the type 316 stainless steel, we evaluated the degree of fatigue damage of the actual material. Furthermore, to verify the evaluation, microstructures of the actual material were investigated with TEM (transmission electron microscope) to observe dislocation densities. As a result, a change in the S-parameter of the actual material from standard as-received material (type 316 stainless steel) was in the range from -0.0013 to 0.0014, while variations in the S-parameter of the standard as-received material were about {+-}0.002, and hence the differences between the actual material and the as-received material were negligible. Moreover, the dislocation density of the actual plant material observed with TEM was almost the same as that of the as-received one. In conclusion, we could confirm the applicability of the positron annihilation lineshape analysis to fatigue damage evaluation of stainless steel. (author)

  2. The development of monitoring techniques for thermal stratification in nuclear plant piping

    International Nuclear Information System (INIS)

    Sim, Cheul Muu; Joo, Young Sang; Yoon, Kwang Sik; Park, Chi Seung; Choi, Ha Lim; Moon, Jae Wha; Bae, Sang Ho.

    1996-12-01

    The conventional nondestructive testing has been performed in those area which are susceptible to thermal stress in according to NRC 88-08,11. In addition to that, it is necessary to set up a monitoring system to prevent severe thermal stress to pipes in early stages and to develop the non-intrusive techniques to diagnose the check valve because the thermal stratification has been caused by the malfunction of the check valve in ECCS pipe. Thermal stratification monitoring system has been designed and installed at ECCS line permanently and surge line temporally in YG nuclear power plant. The data is acceptable in according to TASCS guide line. Also, the data originated from ISMS is useful for the arrangement of a special UT program and stress analysis. Applying a togetherness of acoustics and magnetics signal, it is possible to determine the parameters of the function of the check valve internals without disassembling it. This series of tests show that the accelerometers can be use d to measure and to differentiate the three types of impacts; metal to metal impacts mechanical rubs, and worn internal parts. The magnet sensors can be used to detect the opening/closing of stainless check and fluttering of disk. (author). 50 refs., 5 tabs., 28 figs

  3. Evaluation of vibration and vibration fatigue life for small bore pipe in nuclear power plants

    International Nuclear Information System (INIS)

    Wang Zhaoxi; Xue Fei; Gong Mingxiang; Ti Wenxin; Lin Lei; Liu Peng

    2011-01-01

    The assessment method of the steady state vibration and vibration fatigue life of the small bore pipe in the supporting system of the nuclear power plants is proposed according to the ASME-OM3 and EDF evaluation methods. The GGR supporting pipe system vibration is evaluated with this method. The evaluation process includes the filtration of inborn sensitivity, visual inspection, vibration tests, allowable vibration effective velocity calculation and vibration stress calculation. With the allowable vibration effective velocity calculated and the vibration velocity calculated according to the acceleration data tested, the filtrations are performed. The vibration stress at the welding coat is calculated with the spectrum method and compared with the allowable value. The response of the stress is calculated with the transient dynamic method, with which the fatigue life is evaluated with the Miners linear accumulation model. The vibration stress calculated with the spectrum method exceeds the allowable value, while the fatigue life calculated from the transient dynamic method is larger than the designed life with a big safety margin. (authors)

  4. The development of monitoring techniques for thermal stratification in nuclear plant piping

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Cheul Muu; Joo, Young Sang; Yoon, Kwang Sik; Park, Chi Seung; Choi, Ha Lim; Moon, Jae Wha; Bae, Sang Ho

    1996-12-01

    The conventional nondestructive testing has been performed in those area which are susceptible to thermal stress in according to NRC 88-08,11. In addition to that, it is necessary to set up a monitoring system to prevent severe thermal stress to pipes in early stages and to develop the non-intrusive techniques to diagnose the check valve because the thermal stratification has been caused by the malfunction of the check valve in ECCS pipe. Thermal stratification monitoring system has been designed and installed at ECCS line permanently and surge line temporally in YG nuclear power plant. The data is acceptable in according to TASCS guide line. Also, the data originated from ISMS is useful for the arrangement of a special UT program and stress analysis. Applying a togetherness of acoustics and magnetics signal, it is possible to determine the parameters of the function of the check valve internals without disassembling it. This series of tests show that the accelerometers can be use d to measure and to differentiate the three types of impacts; metal to metal impacts mechanical rubs, and worn internal parts. The magnet sensors can be used to detect the opening/closing of stainless check and fluttering of disk. (author). 50 refs., 5 tabs., 28 figs.

  5. Application of leak-before-break to primary loop piping to eliminate pipe whip restraints in a Spanish nuclear power plant

    International Nuclear Information System (INIS)

    Rodriguez, M.; Esteban, A.

    1990-01-01

    The Spanish plant described in this study is a 982 MWe PWR with a three-loop primary circuit of piping made from centrifugally-cast stainless steel SA351 CF8A. The licensee requested from Consejo de Seguridad Nuclear (CSN) an exemption from the general design criterion, GDC-4, so as to avoid the need to postulate a guillotine rupture of the primary loop piping. The request was based on the generic work performed for a US PWR plant group in order to have such an exemption. As the piping material in the Spanish plant is different from that in the plants included in the generic work, CSN performed a review of the applicability of the generic results to the Spanish plant. Also, aspects such as fatigue evaluation, net section collapse, crack growth and leak detection, specifically analyzed for the Spanish plant, were reviewed. CSN found that fracture toughness test results from generic work are applicable to the Spanish plant; sufficient margin exists against unstable crack extension, and adequate leak detection capability exists with the leakage detection systems available in the plant. Exemption from GDC-4 was approved and CSN authorized the licensee to remove protection devices against dynamic loads from guillotine breaks in the primary coolant loops. (author)

  6. Development of Inspection Technique for Socket Weld of Small Bore Piping in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Yoon, Byungsik; Kim, Yongsik; Lee, Jeongseok

    2013-01-01

    The losses incurred by unplanned shutdowns are significant; consequently, early crack initiation and crack detection, including the detection of fillet weld manufacturing defects, is of the utmost importance. Current inspection techniques are not capable of reliably inspecting socket welds; therefore, new approaches are needed. The new technique must be sensitive to socket weld cracking, which usually initiates from the root, in order to detect the cracking during the early failure phase. In 2008, Kori unit 3 experienced leakage from the drain line socket weld of a steam generator. From this experience, KHNP enforced a management program to focus on enhancing the reliability of small bore socket weld piping inspections. Currently, conventional manual ultrasonic inspection techniques are used to detect service induced fatigue cracks. But there was uncertainty on manual ultrasonic inspection because of limited access to the welds and difficulties with contact between the ultrasonic probe and the OD surface of small bore piping. In this study, phased array ultrasonic inspection techniques are applied to increase inspection speed and reliability. Additionally a manually encoded scanner has been developed to enhance contact conditions and maintain constant signal quality. A phased array UT technique and system was developed to inspect small bore socket welds. The experimental results show all artificial flaws in the specimen were detected and measured. These experimental results show, that the newly developed inspection system, has improved the reliability and speed of small bore socket weld inspection. Based on these results, future works shall focus on additional experiments, with more realistic flaw responses. By applying this technique to the field, we expect that it can improve the integrity of small bore piping in nuclear power plants

  7. Seismic analysis of nuclear piping system

    International Nuclear Information System (INIS)

    Shrivastava, S.K.; Pillai, K.R.V.; Nandakumar, S.

    1975-01-01

    To illustrate seismic analysis of nuclear power plant piping, a simple piping system running between two floors of the reactor building is assumed. Reactor building floor response is derived from time-history method. El Centre earthquake (1940) accelerogram is used for time-history analysis. The piping system is analysed as multimass lumped system. Behaviour of the pipe during the said earthquake is discussed. (author)

  8. Protection Performance Simulation of Coal Tar-Coated Pipes Buried in a Domestic Nuclear Power Plant Using Cathodic Protection and FEM Method

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Y.; Lim, B. T.; Kim, K. S.; Kim, J. W.; Park, H. B. [KEPCO Engineering and Construction Company, Gimcheon (Korea, Republic of); Kim, Y. S.; Kim, K. T. [Andong National University, Andong (Korea, Republic of)

    2017-06-15

    Coal tar-coated pipes buried in a domestic nuclear power plant have operated under the cathodic protection. This work conducted the simulation of the coating performance of these pipes using a FEM method. The pipes, being ductile cast iron have been suffered under considerably high cathodic protection condition beyond the appropriate condition. However, cathodic potential measured at the site revealed non-protected status. Converting from 3D CAD data of the power plant to appropriate type for a FEM simulation was conducted and cathodic potential under the applied voltage and current was calculated using primary and secondary current distribution and physical conditions. FEM simulation for coal tar-coated pipe without defects revealed over-protection condition if the pipes were well-coated. However, the simulation for coal tar-coated pipes with many defects predict that the coated pipes may be severely degraded. Therefore, for high risk pipes, direct examination and repair or renewal of pipes are strongly recommended.

  9. Procedure Development and Qualification of the Phased Array Ultrasonic Testing for the Nuclear Power Plant Piping Weld

    International Nuclear Information System (INIS)

    Yoon, Byung Sik; Yang, Seung Han; Kim, Yong Sik; Lee, Hee Jong

    2010-01-01

    The manual ultrasonic examination for the nuclear power plant piping welds has been demonstrated by using KPD(Korean Performance Demonstration) generic procedure. For automated ultrasonic examination, there is no generic procedure and it should be qualified by using applicable automated equipment. Until now, most of qualified procedures used pulse-echo technique and there is no qualified procedure using phased array technique. In this study, data acquisition and analysis software were developed and phased-array transducer and wedge were designed to implement phased array technique for nuclear power plant in-service inspection. The developed procedure are qualified for performance demonstration for the flaw detection, length sizing and depth sizing. The qualified procedure will be applied for the field examination in the nuclear power plant piping weld inspection

  10. Analysis of Pipe Wall-thinning Caused by Water Chemistry Change in Secondary System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hun; Hwang, Kyeongmo [KEPCO E and C, Gimcheon (Korea, Republic of); Moon, Seung-Jae [Hanyang University, Seoul (Korea, Republic of)

    2015-12-15

    Pipe wall-thinning by flow-accelerated corrosion (FAC) is a significant and costly damage of secondary system piping in nuclear power plants (NPPs). All NPPs have their management programs to ensure pipe integrity from wall-thinning. This study analyzed the pipe wall-thinning caused by changing the amine, which is used for adjusting the water chemistry in the secondary system of NPPs. The pH change was analyzed according to the addition of amine. Then, the wear rate calculated in two different amines was compared at the steam cycle in NPPs. As a result, increasing the pH at operating temperature (Hot pH) can reduce the rate of FAC damage significantly. Wall-thinning is affected by amine characteristics depending on temperature and quality of water.

  11. Application of risk-informed methods to in-service piping inspection in Framatome type nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Jin Hoi; Lee, Jeong Seok; Yun, Eun Sub

    2014-01-01

    The Pressurized water reactor owners group (PWROG) developed and applied a risk-informed in-service inspection (RI-ISI) program, as an alternative to the existing ASME Section XI sampling inspection method. The RI-ISI programs enhance overall safety by focusing inspections of piping at high safety significance (HSS) locations where failure mechanisms are likely to be present. Additionally, the RI-ISI program can reduce nondestructive evaluation (NDE) exams, man-rem exposure for inspectors, and inspection time, among other benefits. The RI-ISI method of in-service piping inspection was applied to 3 units (KSNPs: Korea standard nuclear power plants) and is being deployed to the other units. In this paper, the results of RI-ISI for a Framatome type (France CPI) nuclear power plant are presented. It was concluded that application of RI-ISI to the plant could enhance and maintain plant safety, as well as provide the benefits of greater reliability.

  12. Radioactive recontamination on mechanically polished piping at Shimane-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Umeda, K.; Komoto, I.; Imamura, K.; Kataoka, I.; Uchida, S.

    1998-01-01

    In a series of preventive maintenance tasks for an aging plant, recirculation pipes of Shimane-1 NPP have been replaced by newly fabricated type 316 NG stainless steel pipes. Suppression of shutdown dose rate caused by 60 Co recontamination on the newly replaced piping was one of the major concerns in the recirculation pipe replacement. In order to suppress the shutdown dose rate, control of the 60 Co deposition rate coefficient as well as 60 Co radioactivity in the reactor water are essential. The deposition rate coefficient depends on surface roughness. The coefficient is suppressed by reduction of the effective surface area of pipes through mechanical polishing. Then the inner surface of the pipes was polished mechanically to reduce roughness prior to application in the plant. After measuring and evaluating radioactive recontamination, it was estimated that deposited amounts of radioactive corrosion products on the pipe inner surface would reach the saturated value in a few years, and would not exceed the level before replacement unless water chemistry is degraded. (author)

  13. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J. [Nuclear Res. Inst., Rez (Czech Republic)

    1998-11-01

    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  14. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    International Nuclear Information System (INIS)

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J.

    1998-01-01

    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  15. Nuclear piping and pipe support design and operability relating to loadings and small bore piping

    International Nuclear Information System (INIS)

    Stout, D.H.; Tubbs, J.M.; Callaway, W.O.; Tang, H.T.; Van Duyne, D.A.

    1994-01-01

    The present nuclear piping system design practices for loadings, multiple support design and small bore piping evaluation are overly conservative. The paper discusses the results developed for realistic definitions of loadings and loading combinations with methodology for combining loads under various conditions for supports and multiple support design. The paper also discusses a simplified method developed for performing deadweight and thermal evaluations of small bore piping systems. Although the simplified method is oriented towards the qualification of piping in older plants, this approach is applicable to plants designed to any edition of the ASME Section III or B31.1 piping codes

  16. Test and evaluation about damping characteristics of hanger supports for nuclear power plant piping systems (Seismic Damping Ratio Evaluation Program)

    International Nuclear Information System (INIS)

    Shibata, H.; Ito, A.; Tanaka, K.; Niino, T.; Gotoh, N.

    1981-01-01

    Generally, damping phenomena of structures and equipments is caused by very complex energy dissipation. Especially, as piping systems are composed of many components, it is very difficult to evaluate damping characteristics of its system theoretically. On the other hand, the damping value for aseismic design of nuclear power plants is very important design factor to decide seismic response loads of structures, equipments and piping systems. The very extensive studies titled SDREP (Seismic Damping Ratio Evaluation Program) were performed to establish proper damping values for seismic design of piping as a joint work among a university, electric companies and plant makers. In SDREP, various systematic vibration tests were conducted to investigate factors which may contribute to damping characteristics of piping systems and to supplement the data of the pre-operating tests. This study is related to the component damping characteristics tests of that program. The object of this study is to clarify damping characteristics and mechanism of hanger supports used in piping systems, and to establish the evaluation technique of dispersing energy at hanger support points and its effect to the total damping ability of piping system. (orig./WL)

  17. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

    Slagis, G.C.; Moore, S.E.

    1994-01-01

    Modern-day nuclear plant piping systems are designed with a large number of seismic supports and snubbers that may be detrimental to plant reliability. Experimental tests have demonstrated the inherent ruggedness of ductile steel piping for seismic loading. Present methods to predict seismic loads on piping are based on linear-elastic analysis methods with low damping. These methods overpredict the seismic response of ductile steel pipe. Section III of the ASME Boiler and Pressure Vessel Code stresses limits for piping systems that are based on considerations of static loads and hence are overly conservative. Appropriate stress limits for seismic loads on piping should be incorporated into the code to allow more flexible piping designs. The existing requirements and methods for seismic design of piping systems, including inherent conservations, are explained to provide a technical foundation for modifications to those requirements. 30 refs., 5 figs., 3 tabs

  18. Development of a Multi-Channel Ultrasonic Testing System for Automated Ultrasonic Pipe Inspection of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lee, Hee Jong; Cho, Chan Hee; Cho, Hyun Joon

    2009-01-01

    Currently almost all in-service-inspection techniques, applied in domestic nuclear power plants, are partial to field inspection technique. These kinds of techniques are related to managing nuclear power plants by the operation of foreign-produced inspection devices. There have been so many needs for development of native in-service-inspection device because there is no native diagnosis device for nuclear power plant inspection yet in Korea. In this research, we developed several core techniques to make an automated ultrasonic pipe inspection system for nuclear power plants. A high performance multi-channel ultrasonic pulser/receiver module, an A/D converter module and a digital main CPU module were developed and the performance of the developed modules was verified. The S/N ratio, noise level and signal acquisition performance of the developed modules showed proper level as we designed in the beginning.

  19. Defects in pipe supports attached to concrete structures at Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Nykaenen, J.E.; Reponen, H.; Suominen, J.

    1981-01-01

    The Installation defects in expansion anchors of pipe supports detected in Sweden attracted the attention of the Finnish nuclear authority, the IRP in the autumm of 1979. No serious deficiencies were found at TVO I and II units where expansion anchors tightened by torguing are used. Preliminary inspections at Lo 2 construction plant revealed a great number of defectively installed expansion anchors of the type which is tightened by hitting. This resulted in placing Lo 1 unit into a cold shut-down condition for inspections and reparation. The shut-down lasted for three weeks during which time, in rooms inaccessible during operation, about 2000 expansion anchors were checked and 95% of them repaired or modified. After running up Lo 1 unit, the work continued in accessible rooms and at Lo2 unit for several months. The main fault was the failure to hit the anchor wedge deep enough into its cylinder. A contributing factor may have been too small hole diameters. The anchor tensile tests conducted at the site proved that the insufficient penetration of the wedge drastically reduces the load capacity. (orig./GL)

  20. Application of laser cladding method to small-diameter stainless steel pipes in actual nuclear plant

    International Nuclear Information System (INIS)

    Atago, Y.; Yamadera, M.; Tsuji, H.; Shiraiwa, T.; Kanno, M.

    1995-01-01

    Recently, to prevent stress corrosion cracking (SCC) the material of stainless steel (Type 304), a laser cladding method which produces a highly corrosion-resisting coating (cladding) to be formed on the surface of the material was developed. This is applicable to a long distance and narrow space, because of the good accessibility of the YAG (Yttrium-Aluminum Garnet) laser beam that can be transmitted through an optical fiber. In this method, a paste mixed metallic powder and heating resistive organic solvent is firstly placed on the inner surface of a small pipe and then a YAG laser beam transmitted through an optical fiber is irradiated to the paste, which will be melted and formed a clad subsequently, which is excellent in corrosion resistance. Finally, it can be achieved further resistance against the SCC due to the clad layer formed thus on the surface of the material. Recently, this Laser Cladding method was practically and successfully applied to the actual BWR Nuclear Power Plant in Japan. This report introduces the laser cladding technique, the equipments developed for practical application in the field

  1. A Multi-State Physics Modeling approach for the reliability assessment of Nuclear Power Plants piping systems

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Colli, Davide; Zio, Enrico; Tao, Liu; Tong, Jiejuan

    2015-01-01

    Highlights: • We model piping systems degradation of Nuclear Power Plants under uncertainty. • We use Multi-State Physics Modeling (MSPM) to describe a continuous degradation process. • We propose a Monte Carlo (MC) method for calculating time-dependent transition rates. • We apply MSPM to a piping system undergoing thermal fatigue. - Abstract: A Multi-State Physics Modeling (MSPM) approach is here proposed for degradation modeling and failure probability quantification of Nuclear Power Plants (NPPs) piping systems. This approach integrates multi-state modeling to describe the degradation process by transitions among discrete states (e.g., no damage, micro-crack, flaw, rupture, etc.), with physics modeling by (physic) equations to describe the continuous degradation process within the states. We propose a Monte Carlo (MC) simulation method for the evaluation of the time-dependent transition rates between the states of the MSPM. Accountancy is given for the uncertainty in the parameters and external factors influencing the degradation process. The proposed modeling approach is applied to a benchmark problem of a piping system of a Pressurized Water Reactor (PWR) undergoing thermal fatigue. The results are compared with those obtained by a continuous-time homogeneous Markov Chain Model

  2. Small bore pipe acceptance criteria for watts bar nuclear plant Tennessee Valley Authority

    International Nuclear Information System (INIS)

    Sun, W.S.; Lee, R.L.; Kalyanan, N.

    1991-01-01

    Small bore pipe (≤2 inches NPS) is traditionally analyzed by simplified techniques using Cook Book approach, which yield conservative results. However, reconciliation of these systems for as-built condition where the original criteria is observed to have been exceeded (or due to additions etc.) generally becomes a time consuming and expensive operation since a rigorous computer aided analysis or a detailed hand calculation becomes necessary. The acceptance criteria in this paper can be effectively used in such cases. The approach involves utilizing basic engineering principles and plant specific parameters (such as earthquake spectra) to estimate the system response such as pipe stress due to various loading conditions, piping frequency, support and anchor loads, valve acceleration etc

  3. Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

    International Nuclear Information System (INIS)

    Lydell, Bengt; Olsson, Anders

    2008-01-01

    This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning meetings that were

  4. Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, Bengt (Scandpower Risk Management Inc., Houston, TX (US)); Olsson, Anders (Relcon Scandpower AB, Stockholm (SE))

    2008-01-15

    This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning

  5. Resolving piping analysis issues to minimize impact on installation activities during refueling outage at nuclear power plants

    International Nuclear Information System (INIS)

    Bhavnani, D.

    1996-01-01

    While it is required to maintain piping code compliance for all phases of installation activities during outages at a nuclear plant, it is equally essential to reduce challenges to the installation personnel on how plant modification work should be performed. Plant betterment activities that incorporate proposed design changes are continually implemented during the outages. Supporting analysis are performed to back these activities for operable systems. The goal is to reduce engineering and craft man-hours and minimize outage time. This paper outlines how plant modification process can be streamlined to facilitate construction teams to do their tasks that involve safety related piping. In this manner, installation can proceed by minimizing on the spot analytical effort and reduce downtime to support the proposed modifications. Examples are provided that permit performance of installation work in any sequence. Piping and hangers including the branch lines are prequalified and determined operable. The system is up front analyzed for all possible scenarios. The modification instructions in the work packages is flexible enough to permit any possible installation sequence. The benefit to this approach is large enough in the sense that valuable outage time is not extended and on site analytical work is not required

  6. Assessment and management of ageing of major nuclear power plant components important to safety. Primary piping in PWRs

    International Nuclear Information System (INIS)

    2003-07-01

    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 therefore be 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, 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 to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. The

  7. Assessment of value-impact associated with the elimination of postulated pipe ruptures from the design basis for nuclear power plants

    International Nuclear Information System (INIS)

    Holman, G.S.; Chou, C.K.

    1985-01-01

    The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping systems is also assessed in a less quantitative manner

  8. Development of an on-line ultrasonic system to monitor flow-accelerated corrosion of piping in nuclear power plants

    International Nuclear Information System (INIS)

    Lee, N.Y.; Bahn, C.B.; Lee, S.G.; Kim, J.H.; Hwang, I.S.; Lee, J.H.; Kim, J.T.; Luk, V.

    2004-01-01

    Designs of contemporary nuclear power plants (NPPs) are concentrated on improving plant life as well as safety. As the nuclear industry prepares for continued operation beyond the design lifetime of existing NPP, aging management through advanced monitoring is called for. Therefore, we suggested two approaches to develop the on-line piping monitoring system. Piping located in some position is reported to go through flow accelerated corrosion (FAC). One is to monitor electrochemical parameters, ECP and pH, which can show occurrence of corrosion. The other is to monitor mechanical parameters, displacement and acceleration. These parameters are shown to change with thickness. Both measured parameters will be combined to quantify the amount of FAC of a target piping. In this paper, we report the progress of a multidisciplinary effort on monitoring of flow-induced vibration, which changes with reducing thickness. Vibration characteristics are measured using accelerometers, capacitive sensor and fiber optic sensors. To theoretically support the measurement, we analyzed the vibration mode change in a given thickness with the aid of finite element analysis assuming FAC phenomenon is represented only as thickness change. A high temperature flow loop has been developed to simulate the NPP secondary condition to show the applicability of new sensors. Ultrasonic transducer is introduced as validation purpose by directly measuring thickness. By this process, we identify performance and applicability of chosen sensors and also obtain base data for analyzing measured value in unknown conditions. (orig.)

  9. Development of testing system for the thermo-mechanical fatigue crack analysis of nuclear power plant pipes

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Kim, Maan Won; Lee, Bong Sang

    2003-12-01

    Fatigue crack growth analysis plays an important role in the structural integrity assessment or the service life calculation of the nuclear power plant pipes. To obtain the material properties as a basic data to achieve an accurate crack growth analysis, a lot of tests and numerical crack growth simulations have been done for decades. The BS 7910 or the ASME Boiler and Pressure Vessel Code Section XI, generally used to evaluate crack growth behavior, were made under the based on simple stress states or at the evaluated isothermal temperature. It is well known that the ASME code could sometimes give so conservative results in some cases of which the cracked components are experiencing with cyclic thermal shock. In this report, we suggested a method for the life assessment of a crack embedded in nuclear power plant pipes under the thermal-mechanical fatigue loads. We here use the numerical method to get the temperature history for thermal- mechanical fatigue crack growth test. And then we can calculate the remaining life time of the pipe by using the fracture mechanics and the test results together. For this purpose, we constructed a thermal-mechanical fatigue crack growth testing system. We also gave a lot of review about recent researches in the experimental field of thermal-mechanical fatigue analysis

  10. Stochastic evaluation of the dynamic response and the cumulative damage of nuclear power plant piping

    International Nuclear Information System (INIS)

    Suzuki, Kohei; Aoki, Shigeru; Hanaoka, Masaaki

    1981-01-01

    This report deals with a fundamental study concerning an evaluation of uncertainties of the nuclear piping response and cumulative damage under excess-earthquake loadings. The main purposes of this study cover following several problems. (1) Experimental estimation analysis of the uncertainties concerning the dynamic response and the cumulative failure by using piping test model. (2) Numerical simulation analysis by Monte Carlo method under the assumption that relation between restoring force and deformation is characterized by perfectly elasto-plastic one. (Checking the mathematical model.) (3) Development of the conventional uncertainty estimating method by introducing a perturbation technique based on an appropriate equivalently linearized approach. (Checking the estimation technique.) (4) An application of this method to more realistical cases. Through above mentioned procedures some important results are obtained as follows; First, fundamental statistical properties of the natural frequencies and the number of cycle to failure crack initiation are evaluated. Second, the effect of the frequency fluctuation and the yielding fluctuation are estimated and examined through Monte Carlo simulation technique. It has become clear that the yielding fluctuation gives significant effect on the piping power response up to its failure initiation. Finally some results through proposed perturbation technique are discussed. Statistical properties estimated coincide fairly well with those through numerical simulation. (author)

  11. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  12. Flow Accelerated Erosion-Corrosion (FAC) considerations for secondary side piping in the AP1000{sup R} nuclear power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Vanderhoff, J. F.; Rao, G. V. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Stein, A. [Shaw Power Nuclear, 1000 Technology Center Drive, Stoughton, MA 02072 (United States)

    2012-07-01

    The issue of Flow Accelerated Erosion-Corrosion (FAC) in power plant piping is a known phenomenon that has resulted in material replacements and plant accidents in operating power plants. Therefore, it is important for FAC resistance to be considered in the design of new nuclear power plants. This paper describes the design considerations related to FAC that were used to develop a safe and robust AP1000{sup R} plant secondary side piping design. The primary FAC influencing factors include: - Fluid Temperature - Pipe Geometry/layout - Fluid Chemistry - Fluid Velocity - Pipe Material Composition - Moisture Content (in steam lines) Due to the unknowns related to the relative impact of the influencing factors and the complexities of the interactions between these factors, it is difficult to accurately predict the expected wear rate in a given piping segment in a new plant. This paper provides: - a description of FAC and the factors that influence the FAC degradation rate, - an assessment of the level of FAC resistance of AP1000{sup R} secondary side system piping, - an explanation of options to increase FAC resistance and associated benefits/cost, - discussion of development of a tool for predicting FAC degradation rate in new nuclear power plants. (authors)

  13. Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O., E-mail: vasconv@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2016-07-01

    During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

  14. Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

    International Nuclear Information System (INIS)

    Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O.

    2016-01-01

    During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

  15. Thin-plate-type embedded ultrasonic transducer based on magnetostriction for the thickness monitoring of the secondary piping system of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Tae Hoon; Cho, Seung Hyun [Center for Safety Measurement, Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

    2016-12-15

    Pipe wall thinning in the secondary piping system of a nuclear power plant is currently a major problem that typically affects the safety and reliability of the nuclear power plant directly. Regular in-service inspections are carried out to manage the piping system only during the overhaul. Online thickness monitoring is necessary to avoid abrupt breakage due to wall thinning. To this end, a transducer that can withstand a high-temperature environment and should be installed under the insulation layer. We propose a thin plate type of embedded ultrasonic transducer based on magnetostriction. The transducer was designed and fabricated to measure the thickness of a pipe under a high-temperature condition. A number of experimental results confirmed the validity of the present transducer.

  16. A study of the long-range inspection method for on-line monitoring of pipes in nuclear power plants

    International Nuclear Information System (INIS)

    Eom, Heung Seop; Lim, Sa Hoe; Kim, Jae Hee; Kim, Young H.; Song, Sung Jin

    2005-01-01

    Deployment of an advanced on-line monitoring of the component integrity offers the prospect of an improved performance, enhanced safety, and reduced overall cost for nuclear power plants (NPPs). Also ultrasonic guided ultrasonic wave has been known as one of the promising techniques that could be utilized for on-line monitoring, because it enables us to undertake a long-range inspection of structures such as plates and pipes. The present work is aimed at developing a new method using ultrasonic guided waves for the on-line monitoring of pipes. For this purpose we fabricated the necessary hardware and carried out transmitter tuning, group velocity measurement, receiver tuning, and mode identification. Finally we carried out an experiment on a long-range inspection with the developed hardware and the techniques. In the experiment, we could detect the flaws at a distance of about 20M from the transmitter, and we could verify the possibility of using the developed hardware and techniques for on-line monitoring of pipes in NPPs

  17. Material property requirements for application leak-before-break technology on nuclear power plant high-energy piping

    International Nuclear Information System (INIS)

    Li Chengliang; Deng Xiaoyun; Yin Zhiying; Liu Meng

    2012-01-01

    The application of leak-before-break (LBB) technology on nuclear power plant high-energy piping systems can improve their safety and economy, while propose some new requirements on testing material properties. The U.S. Nuclear Regulatory Commission's LBB related standard review plan and implementation specifications were analyzed, and test items, object, temperature, quantity and thermal aging effect of five general requirements were summarized. In addition, four key testing technical requirements, such as specimen size, side grooves, strain range and the orientation of specimens were also discussed to ensure the test data usefulness, representativeness and integrity. This study can provide some guidance for the aforementioned test program on domestic materials. (authors)

  18. Development of an on-site measurement method for residual stress in primary system piping of nuclear power plants

    International Nuclear Information System (INIS)

    Maekawa, Akira; Takahashi, Shigeru; Fujiwara, Masaharu

    2014-01-01

    In residual stress measurement for large-scale pipes and vessels in high radiation areas and highly contaminated areas of nuclear plants, it is difficult to bring the radioactivated pipes and vessels out of the areas as they are. If they can brought out, it is very burdensome to handle them for the measurement. Development of an on-site measurement method of residual stress which can be quickly applied and has sufficient measurement accuracy is desirable. In this study, a new method combining an electric discharge skim-cut method with a microscopic strain measurement method using markers was proposed to realize the on-site residual stress measurement on pipes in high radiation areas and highly contaminated areas. In the electric discharge skim-cut method, a boat-type sample is skimmed out of a pipe outer/inner surface using electric discharge machining and released residual stress is measured. The on-site measurement of residual stress by the method can be done using a small, portable electric discharge machine. In the microscopic strain measurement method using markers, the residual stress is estimated by microscopic measurement of the distance between markers after the stress release. The combination of both methods can evaluate the residual stress with the same accuracy as conventional methods offer and it can achieve reduction of radiation exposure in the measurement because the work is done simply and rapidly. In this study, the applicability of the electric discharge skim-cut method was investigated because the applicability of the microscopic strain measurement method using markers was confirmed previously. The experimental examination clarified the applicable conditions for the residual stress measurement with the same accuracy as the conventional methods. Furthermore, the electric discharge machining conditions using pure water as the machining liquid was found to eliminate the amount of liquid radioactive waste completely. (author)

  19. Leak detection in the primary reactor coolant piping of nuclear power plant by applying beam-microphone technology

    International Nuclear Information System (INIS)

    Kasai, Yoshimitsu; Shimanskiy, Sergey; Naoi, Yosuke; Kanazawa, Junichi

    2004-01-01

    A microphone leak detection method was applied to the inlet piping of the ATR-prototype reactor, Fugen. Statistical analysis results showed that the cross-correlation method provided the effective results for detection of a small leakage. However, such a technique has limited application due to significant distortion of the signals on the reactor site. As one of the alternative methods, the beam-microphone provides necessary spatial selectivity and its performance is less affected by signal distortion. A prototype of the beam-microphone was developed and then tested at the O-arai Engineering Center of the Japan Nuclear Cycle Development Institute (JNC). On-site testing of the beam-microphone was carried out in the inlet piping room of an RBMK reactor of the Leningrad Nuclear Power Plant (LNPP) in Russia. A leak sound imitator was used to simulate the leakage sound under the leakage flow condition of 1-3 gpm (0.23-0.7 m 3 /h). Analysis showed that signal distortion does not seriously affect the performance of this method, and that sound reflection may result in the appearance of ghost sound sources. The test results showed that the influences of sound reflection and background noise were smaller at the high frequencies where the leakage location could be estimated with an angular accuracy of 5deg which is the range of localization accuracy required for the leak detection system. (author)

  20. Leak-before-break analysis of a dissimilar metal welded joint for connecting pipe-nozzle in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Gong, N. [MOE Key Laboratory of Pressurized System and Safety, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China); Wang, G.Z., E-mail: gzwang@ecust.edu.cn [MOE Key Laboratory of Pressurized System and Safety, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China); Xuan, F.Z.; Tu, S.T. [MOE Key Laboratory of Pressurized System and Safety, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China)

    2013-02-15

    Highlights: ► Leak-before-break (LBB) analysis for a dissimilar metal weld joint (DMWJ) is made. ► Pipe-nozzle geometry and inhomogeneous material property of DMWJ are incorporated. ► LBB behavior of a defect can be assessed by LBB assessment diagram and LBB curve. ► Feasibility region of LBB is enlarged with decreasing load and increasing J{sub R}. -- Abstract: This paper presents a leak-before-break (LBB) analysis for a dissimilar metal welded joint (DMWJ) connected the safe end to pipe-nozzle of a reactor pressure vessel of which is relevant to safety of nuclear power plant. Three-dimensional finite element analysis models were built for the DMWJ structure, and the initial inner circumferential surface cracks were postulated at the interface between A508 steel and buttering Alloy82. Based on the elastic–plastic fracture mechanics theory of J-integral, the crack growth stability was analyzed, and the pipe-nozzle geometry effect and inhomogeneous material properties of the DMWJ have been incorporated. Base on the analysis results, the LBB curves and LBB assessment diagrams were constructed for the DMWJ, and effects of applied bending moment loads and J-resistance curves of materials on LBB behavior were analyzed. The results show that the LBB behavior of a defect in the DMWJ under an upmost severe load can be assessed and predicted by plotting the defect size and its propagation path in the LBB assessment diagrams. With decreasing the maximum bending moment load and increasing the crack growth resistance of materials, the ligament instability lines shift upward and the critical crack length lines move to the right in the LBB assessment diagrams, which leads to enlargement of the feasibility region in the LBB behavior.

  1. An overview of environmental degradation of materials in nuclear power plant piping systems

    International Nuclear Information System (INIS)

    Shack, W.J.

    1987-08-01

    Piping in light water reactor (LWR) power systems is affected by several types of environmental degradation: intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs) has required research, inspection, and mitigation programs that will ultimately cost several billion dollars; erosion-corrosion of carbon steel piping has been observed frequently in the secondary systems of both BWRs and pressurized water reactors (PWRs); the effect of the BWR environment can greatly diminish the design margin inherent in the ASME Section III fatigue design curves for carbon steel piping; and cast stainless steels are subject to embrittlement after extended thermal aging at reactor operating temperatures. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

  2. Safety catching device for pipes in missile shielding cylinders of nuclear power plants

    International Nuclear Information System (INIS)

    Hering, S.; Doll, B.

    1976-01-01

    The safety catching device consists of a steel wire passed in U-shape around the pipe to be caught and supported by two anchor ties embedded in the concrete of the missile shielding cylinder. This flexible catching device is to cause the energy released in case of a pipe rupture to be absorbed and no dangerous bending shesses to be transferred to the walls of the missile shielding cylinder. (UWI) [de

  3. An experimental study on damping characteristics of mechanical snubber for nuclear power plant piping systems

    International Nuclear Information System (INIS)

    Chiba, T.; Kobayashi, H.; Kitamura, K.; Ando, K.; Koyanagi, R.

    1983-01-01

    The objectives of this study are 1) to clarify the damping characteristics and the dynamic stiffness of mechanical snubber, 2) to take the damping characteristics of mechanical snubber into the damping evaluation method obtained in SDREP. Therefore, following vibration tests were conducted. 1) Component test: As a first step, mechanical snubbers were excited with sinusoidal wave, and damping ratio and dynamic stiffness were measured at several loading levels. 2) Piping model test: Second, a 8'' diameter x 16 m length 3-dimensional piping model simulating the supporting conditions of actual piping systems was tested. Damping ratio and made shapes of piping model with mechanical snubbers were measured at several supporting conditions and response levels. From the results of these tests, the damping characteristics and the dynamic stiffness of mechanical snubber can be summarized as follows: 1) The damping effect of mechanical snubber is as strong as that of oil snubber. 2) Mechanical snubber contributes effectively to the damping of piping system, and it is indicated that the damping characteristics of mechanical snubber is applicable to the damping evaluation method obtained in SDREP. (orig./HP)

  4. Piping failures in United States nuclear power plants 1961-1995

    International Nuclear Information System (INIS)

    Bush, S.H.; Do, M.J.; Slavich, A.L.; Chockie, A.D.

    1996-01-01

    Over 1500 reported piping failures were identified and summarized based on an extensive review of tens of thousands of event reports that have been submitted to the US regulatory agencies over the last 35 years. The data base contains only piping failures; failures in vessels, pumps, valves and steam generators or any cracks that were not through-wall are not included. It was observed that there has been a marked decrease in the number of failures after 1983 for almost all sizes of pipes. This is likely due to the changes in the reporting requirements at that time and the corrective actions taken by utilities to minimize fatigue failures of small lines and IGSCC in BWRs. One failure mechanism that continues to occur is erosion-corrosion, which accounts for most of the ruptures reported and probably is responsible for the absence of downward trends in ruptures. Fatigue-vibration is also a significant contributor to piping failures. However, most of such events occur in lines approx. one inch or less in diameter. Together, erosion-corrosion and fatigue-vibration account for over 43 per cent of the failures. The overwhelming majority of failures have been leaks, over half the failures occurred in pipes with a diameter of one inch or less. Included in the report is a listing of the number of welds in various systems in LWRs

  5. An experimental study of damping characteristics with emphasis on insulation for nuclear power plant piping system (Seismic Damping Ratio Evaluation Program)

    International Nuclear Information System (INIS)

    Shibata, H.; Ito, M.; Hayashi, T.; Chiba, T.; Kobayashi, H.; Kitamura, K.; Ando, K.; Koyanagi, R.

    1981-01-01

    To clarify the damping characteristics and mechanism in nuclear power plant piping systems, the study group was established and conducted to study SDREP (Seismic Damping Ratio Evaluation Program). As the Phase II of this study, vibration tests were conducted to investigate factors which might contribute to damping characteristics of piping systems. These tests are composed of the next three model tests: 1) The component damping characteristics test of thermal insulator 2) The simplified piping model test 3) The scale model test. In these tests, we studied damping characteristics with emphasis on thermal insulator (mainly calcium silicate insulator). The acceleartion level of pipings is the same as that of the actual seismic response. The excitation was by sinusoidal sweep method using the shaking table and by free vibration method using snapback. (orig./RW)

  6. Safety catching device for pipe lines in missile shielding cylinders of nuclear power plants

    International Nuclear Information System (INIS)

    Hering, S.; Doll, B.

    1975-01-01

    The safety catching device for pipes in the missile shielding cylinders consists of a flexible steel cable surrounding the pipe in a distance in U-shape. The arrester cable - which works as a spring and is freely movable in all directions - is attached to the cylinder wall. For this, the ends of the cable are primarily fastened to anchor boxes which are then inserted in a stay tube with the same axis as the cable ends. The anchor boxes are fastened to the outer wall of the missile shielding cylinder by anchor bolts and holding plates. (DG/AK) [de

  7. Effects of supporting structures on dynamic response of nuclear power plant equipment and piping systems

    International Nuclear Information System (INIS)

    Stoykovich, M.

    1982-01-01

    This paper presents the evaluation of the effects of supporting structures in dynamic analysis of equipment or piping systems, which involves formulations for determining reduced stiffness and mass matrices associated with the number of degrees of freedom corresponding to the support nodal points of a finite element model. Also, evaluation of a composite damping matrix associated with different damping properties of supporting structures, equipment, and piping systems is considered. Determination of spring constants, effective masses and mass moments of inertia, and damping values as fractions of critical damping on the basis of the theory of rigid bases on the surfaces of an elastic halfspace is demonstrated

  8. Development of inspection technology for inner wall pipe of aging nuclear power plant

    International Nuclear Information System (INIS)

    Ito, Fuyumi; Nishimura, Akihiko

    2013-01-01

    Careful inspection should be paid on aging nuclear power plants. Due to the Fukushima BWR accident, more advanced inspection techniques are now requested in Japan. To find SCC along welded sections by Ultrasonic Testing or Eddy Current Testing is difficult due to the low S/N. Here we propose to apply Magnetic particle Testing (MT) on the inspection. MT uses magnetic particles uniting fluorescent pigment. It is a weak point of MT that uniting particles and pigment is breakable. To extend the lifetime, we developed unique capsule for the magnetic particle to coexist with fluorescent pigment. In addition, Laser-Induced Breakdown Spectroscopy (LIBS) used for the laser cleaning of materials, is reported in this paper as a preliminary experiment. The intensity of 621nm peak gradually decreases over time. This result will become a measure of the degree of oxide layer removal. (author)

  9. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

    Energy Technology Data Exchange (ETDEWEB)

    1985-04-01

    This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are addressed. Effects of current regulatory requirements for piping design are evaluated, and recommendations for immediate licensing action, changes in existing requirements, and research programs are presented. Additional background information and suggestions given by consultants are also presented.

  10. Some considerations for establishing seismic design criteria for nuclear plant piping

    International Nuclear Information System (INIS)

    Chen, W.P.; Chokshi, N.C.

    1997-01-01

    The Energy Technology Engineering Center (ETEC) is providing assistance to the U.S. NRC in developing regulatory positions on the seismic analysis of piping. As part of this effort, ETEC previously performed reviews of the ASME Code, Section III piping seismic design criteria as revised by the 1994 Addenda. These revised criteria were based on evaluations by the ASME Special Task Group on Integrated Piping Criteria (STGIPC) and the Technical Core Group (TCG) of the Advanced Reactor Corporation (ARC) of the earlier joint Electric Power Research Institute (EPRI)/NRC Piping ampersand Fitting Dynamic Reliability (PFDR) program. Previous ETEC evaluations reported at the 23rd WRSM of seismic margins associated with the revised criteria are reviewed. These evaluations had concluded, in part, that although margins for the timed PFDR tests appeared acceptable (>2), margins in detuned tests could be unacceptable (<1). This conclusion was based primarily on margin reduction factors (MRFs) developed by the ASME STGIPC and ARC/TCG from realistic analyses of PFDR test 36. This paper reports more recent results including: (1) an approach developed for establishing appropriate seismic margins based on PRA considerations, (2) independent assessments of frequency effects on margins, (3) the development of margins based on failure mode considerations, and (4) the implications of Code Section III rules for Section XI

  11. Piping Stress analysis for primary system of nuclear power plant AP-600

    International Nuclear Information System (INIS)

    Tjahjono, Hendro; Arhatari, B.D.; W, Pustandyo; Sitandung, J.B; Sudarmaji, Djoko

    1999-01-01

    Piping stress analysis for AP-600 primary system has been done using software CAEPIPE and PS-CAEPIPE. The loading applied to the system are static and seismic category I and II piping in reactor building have been analysed, those are PXS-900, CVS-110, PCS-030, CAS-700 and CCS-050. These system contain pipes with the normal diameter of 1 , 2 , 4 a nd 8 . The design pressures are in the range of 150oF to 300oF. The acceleration taken as input in PS-CAEPIPE is based on seismic response spectra of floor the piping is located. In CAEPIPE, the acceleration taken from the peak of response spectra multiplied by 1.7 all of the acceleration in this case are no more than 0.36g. The result shows that after locating some supports, all system are acceptable without snubbers. The maximum stress are 11210 psi for deadweight load and 35593 psi for total load (the allowable values are 15000 psi and 45000 psi). The maximum displacement are 0.123 in for deadweight load, 1.474 in for hot load seismic load (the allowable values are 0.125 in for deadweight and 2.5 in for total load). The difference results of the both software is mainly in seismic calculation where mare parameters can be evaluated by PS-CAEPIPE including to evaluate valves acceleration in seismic condition

  12. Wall thinning of piping in power plants

    International Nuclear Information System (INIS)

    Ohta, Joji; Inada, Fumio; Morita, Ryo; Kawai, Noboru; Yoneda, Kimitoshi

    2005-01-01

    Major mechanisms causing wall thinning of piping in power plants are flow accelerated corrosion (FAC), cavitation erosion and droplet erosion. Their fundamental aspects are reviewed on the basis of literature data. FAC is chemical process and it is affected by hydrodynamic factors, temperature, pH, dissolved oxygen concentration and chemical composition of materials. On the other hand, cavitation erosion and droplet erosion are mechanical process and they are mainly affected by hydrodynamic factors and mechanical properties of materials. Evaluation codes for FAC and mitigation methods of FAC and the erosion are also described. Wall thinning of piping is one of public concerns after an accident of a pipe failure at Mihama Nuclear Power Plant Unit 3, Kansai Electric Power Co., Inc., in August 2004. This paper gives comprehensive understanding of the wall thinning mechanism. (author)

  13. Passive cooling applications for nuclear power plants using pulsating steam-water heat pipes

    International Nuclear Information System (INIS)

    Aparna, J.; Chandraker, D.K.

    2015-01-01

    Gen IV reactors incorporate passive principles in their system design as an important safety philosophy. Passive safety systems use inherent physical phenomena for delivering the desired safe action without any external inputs or intrusion. The accidents in Fukushima have renewed the focus on passive self-manageable systems capable of unattended operation, for long hours even in extended station blackout (SBO) and severe accident conditions. Generally, advanced reactors use water or atmospheric air as their ultimate heat sink and employ passive principles in design for enhanced safety. This paper would be discussing the experimental results on pulsating steam water heat-pipe devices and their applications in passive cooling. (author)

  14. Analysis of gamma ray intensity on the S/C vent pipes area in the unit 2 reactor building of the Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The robot is equipped with cameras, a dosimeter, and 2 DOF (degree of freedom) manipulation arms. It loads a small vehicle equipped with a camera that can access and inspect narrow areas. TEPCO is using the four-legged walking robot to inspect the suppression chamber (S/C) area of the unit 2 reactor building basement in the Fukushima Daiichi Nuclear Power Plant. The robot carried out 6 missions for about four months, from 11 December, 2012 to 15 March, 2013, where it examined an evidence of a leakage of radioactivity contaminated water in the S/C area of unit 2 reactor building. When a camera's signal processing unit, which is consist of ASIC and FPGA devices manufactured by a CMOS fabrication process, is exposed to a higher dose rate gamma ray, the speckle distribution in the camera image increase more. From the inspection videos, released by TEPCO, of the underground 8 vent pipes in the unit 2 reactor building, we analyzed the speckle distribution from the high dose-rate gamma rays. Based on the distribution of the speckle, we attempted to characterize the vent pipe with much radioactivity contaminated materials among the eight vent pipes connected to the PCV. The numbers of speckles viewed in the image of a CCD (or CMOS) camera are related to an intensity of the gamma ray energy emitted by a nuclear fission reaction from radioactivity materials. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. Therefore, calculating the speckles counts, we can determine the vent pipe with relatively most radioactivity-contaminated materials among the other vent pipes. From the comparison of speckles counts calculated in the inspection image of the vent pipe with the speckles counts extracted by gamma ray irradiation experiment of the same small vehicle camera model loaded with the four-legged walking robot, we can qualitatively estimate the gamma ray dose-rate in the S/C vent pipe area of the

  15. Development of acoustic leak detection and localization methods for inlet piping of fugen nuclear power plant

    International Nuclear Information System (INIS)

    Shimanskiy, Sergey; Iijima, Takashi; Naoi, Yosuke

    2004-01-01

    The development work carried out on Fugen NPP is focused on detection of a small leakage on the reactor's inlet feeder pipes at an early stage by an acoustic leak detection method with usage of high-temperature resistant microphones. Specifically, the leak rate of 0.046m 3 /h has been chosen as a target detection capability for this system. A cross-correlation technique has been studied for leak detection under low signal-noise ratios. The study shows that the sound diffusion on piping causes distortion of leak signals that results in their low correlation. A leak-location estimator and multi-channel correlation value, associated with estimated leak position, have been employed to detect such low-correlated leak signals. A method based on cross-correlation of signal spectral components has been proposed to deal with non-stationary leak signals. Joint-Time-Frequency-Analysis has been applied to analyze such signals, whilst a Wavelet decomposition technique has been used to extract their short-term spectral fluctuations. Since the spectral components are less affected by signal distortion, they provide higher correlation value and can be applied for leak detection under lower signal-noise ratios. The possibility of detecting and locating a small leakage by the methods proposed has been demonstrated by a number of simulation tests conducted on the Fugen NPP site. (author)

  16. Efficient improvement of nuclear power plant safety by reorganization of risk-informed safety importance evaluation methods for piping welded portions

    Energy Technology Data Exchange (ETDEWEB)

    Irie, Takashi; Hanafusa, Hidemitsu; Suyama, Takeshi [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Morota, Hidetsugu; Kojima, Sigeo; Mizuno, Yoshinobu [Computer Software Development Co., Ltd., Tokyo (Japan)

    2002-09-01

    In this work, risk information was used to evaluate the safety importance of piping welded portions which were important for plant operation and maintenance of nuclear power plants. There are two types of risk-informed safety importance evaluation methods, namely the ASME method and the EPRI method. Since both methods have advantages and disadvantages, elements of each method were combined and reorganized. Considerations included whether the degradation mechanisms would be objectively evaluated and whether plant safety would be efficiently improved. The most objective and efficient method was as follows. Piping failure potential is quantitatively and objectively evaluated for failure with probabilistic fracture mechanics (PFM) and for other degradation mechanisms with empirical failure rates, and conditional core damage probability (CCDP) is calculated with PSA. This method reduces the inspected segment numbers to 1/4 of the deterministic method and increases the ratio of risk, which is covered by the inspected segments, to total risk from 80% of the deterministic method to 95%. Piping inspection numbers decreased for safety injection systems that were required the inspections by the deterministic method. Piping inspections were required for part of main feed water and main steam systems that were not required the inspections by the deterministic method. (author)

  17. Reserve seismic capacity determination of a nuclear power plant braced frame with piping

    International Nuclear Information System (INIS)

    Nelson, T.A.

    1979-01-01

    The Lawrence Livermore Laboratory has been asked by the U.S. Nuclear Regulatory Commission to investigate the inelastic behavior of a representative non-category I structure to determine the amount of reserve seismic capacity that is available beyond elastic design levels. This reserve capacity can be an important consideration when evaluating the ability of existing structures to withstand upgraded seismic hazards. (orig.)

  18. The construction for remediation work of contaminated water at Fukushima Daiichi Nuclear Power Plant. Closure work of seawater piping trench and screen pump chamber

    International Nuclear Information System (INIS)

    Hibi, Yasuki; Yanai, Shuji; Nishikori, Kazumasa; Soma, Yu

    2016-01-01

    In the seawater piping trench of Fukushima Daiichi Nuclear Power Plant, highly contaminated water was stagnating, which flowed in from the reactor building and turbine building affected by the tsunami caused by the Tohoku Pacific Ocean Earthquake. Although the screen pump chamber, adjacent to the seawater piping trench, escaped from the inflow and retention of contaminated water, it was exposed to the leakage risk of contaminated water from the seawater piping trench. As measures against these conditions, the following emergency work was applied: (1) contaminated water replacement and removal operation based on the implantation of fillers into the seawater piping trench, and (2) closure operation of the screen pump chamber by implanting fillers into the screen pump chamber. In face of these operations, long-distance underwater flow special filler, high workable concrete, and underwater non-separation concrete were developed and used. The implantation of the long-distance underwater-flow special fillers into the seawater piping trench was successfully completed by filling to the tunnel top without gap and without water head difference, and by preventing the occurrence of movement or water path formation of the fillers in the initial curing process. Other fillers were also able to be implanted as planned. The leakage risk of contaminated water to the periphery could be suppressed to a large extent by this work. (A.O.)

  19. Automatic inspection of the main primary circuit piping at the south Ukraine nuclear power plant

    International Nuclear Information System (INIS)

    Tanarro, A.

    1995-01-01

    In response to GOSKOMATOM requests, the TPEG (from The Commission of the European Communities) undertook development of the general approach to the Project, or Terms of Reference, with a view to procuring complete financing of the project through TACIS funds. Subsequently UNESA (Unidad Electrica S.A.), as the organization responsible for the On-site Assistance Project for South Ukraine NPP, jointly reviewd the contents of the project with the plant and coordinated development throughout the initial phase, which finishes in May 1995. From the very start the definition of the Project covered all the Activities required to ensure its final objective: full autonomy for the Plant to perform inspections and the capacity required for this to be achieved to a suitable level of quality

  20. Development of super duplex stainless steel for water-supply pipe and valve in nuclear plants

    International Nuclear Information System (INIS)

    Park, Chan Jin; Kim, Jun Sick; Kwon, Hyuk Sang; Park, Young Hwan; Lee, Zin Hyung

    2000-01-01

    Austenitic-ferritic duplex stainless steels are very attractive as material for water-supply facilities in atomic power plants where both high mechanical strength and excellent resistance to localized and stress corrosion are required. However, these alloys have a problem to get sensitive to embrittlement when exposed to temperatures of 250 ∼ 1050 deg C. So far, there have been large efforts to improve this alloy. In this paper, a new developed alloy designed to improve not resistance to the embrittlement but also mechanical and corrosion properties compared with existing commercial alloys were introduced with some experimental results. (author)

  1. Probabilistic residual life assessment of high temperature pipings in nuclear power plants against creep fatigue damage: final report

    International Nuclear Information System (INIS)

    Gupta, Sanjay

    2014-02-01

    Residual life assessment of components of nuclear power plants is essential for their operational safety, reliability and financial viability. The high risks involved in the event of failures in nuclear power plants have led to the development of design philosophies that incorporate extreme conservatism in design. The implications of such conservatism in design leads to more frequent maintenance operations than necessary

  2. Nuclear class 1 piping stress analysis

    International Nuclear Information System (INIS)

    Lucas, J.C.R.; Maneschy, J.E.; Mariano, L.A.; Tamura, M.

    1981-01-01

    A nuclear class 1 piping stress analysis, according to the ASME code, is presented. The TRHEAT computer code has been used to determine the piping wall thermal gradient. The Nupipe computer code was employed for the piping stress analysis. Computer results were compared with the allowable criteria from the ASME code. (Author) [pt

  3. Quality control of stainless steel pipings for nuclear power generation

    International Nuclear Information System (INIS)

    Miki, Minoru; Kitamura, Ichiro; Ito, Hisao; Sasaki, Ryoichi

    1979-01-01

    The proportion of nuclear power in total power generation is increasing recently in order to avoid the concentrated dependence on petroleum resources, consequently the reliability of operation of nuclear power plants has become important. In order to improve the reliability of plants, the reliability of each machine or equipment must be improved, and for the purpose, the quality control at the time of manufacture is the important factor. The piping systems for BWRs are mostly made of carbon steel, and stainless steel pipings are used for the recirculation system cooling reactors and instrumentation system. Recently, grain boundary type stress corrosion cracking has occurred in the heat-affected zones of welded stainless steel pipings in some BWR plants. In this paper, the quality control of stainless steel pipings is described from the standpoint of preventing stress corrosion cracking in BWR plants. The pipings for nuclear power plants must have sufficient toughness so that the sudden rupture never occurs, and also sufficient corrosion resistance so that corrosion products do not raise the radioactivity level in reactors. The stress corrosion cracking occurred in SUS 304 pipings, the factors affecting the quality of stainless steel pipings, the working method which improves the corrosion resistance and welding control are explained. (Kako, I.)

  4. Effects of Induction Heat Bending and Heat Treatment on the Boric Acid Corrosion of Low Alloy Steel Pipe for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki-Tae; Kim, Young-Sik [Andong National University, Gyeongbuk (Korea, Republic of); Chang, Hyun-Young; Park, Heung-Bae [KEPCO EandC, Gyeongbuk (Korea, Republic of); Sung, Gi-Ho; Shin, Min-Chul [Sungil SIM Co. Ltd, Busan (Korea, Republic of)

    2016-11-15

    In many plants, including nuclear power plants, pipelines are composed of numerous fittings such as elbows. When plants use these fittings, welding points need to be increased, and the number of inspections also then increases. As an alternative to welding, the pipe bending process forms bent pipe by applying strain at low or high temperatures. This work investigates how heat treatment affects on the boric acid corrosion of ASME SA335 Gr. P22 caused by the induction heat bending process. Microstructure analysis and immersion corrosion tests were performed. It was shown that every area of the induction heat bent pipe exhibited a high corrosion rate in the boric acid corrosion test. This behavior was due to the enrichment of phosphorous in the ferrite phase, which occurred during the induction heat bending process. This caused the ferrite phase to act as a corrosion initiation site. However, when re-heat treatment was applied after the bending process, it enhanced corrosion resistance. It was proved that this resistance was closely related to the degree of the phosphorus segregation in the ferrite phase.

  5. Study on filling materials suitable for seawater piping trench closure work at Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    Yanai, Shuji; Hibi, Yasuki; Nishikori, Kazumasa; Sato, Keita

    2016-01-01

    Highly contaminated water leaking from the reactor buildings and turbine buildings damaged by the 2011 Great East Japan Earthquake has accumulated in the seawater piping trenches of Fukushima Daiichi Nuclear Power Station Units 2, 3, and 4. In November 2014, work commenced to replace and remove this contaminated water by filling the trenches with filling materials, and this work was completed in December 2015. This paper summarizes the contents of this study on various filling materials, including special fillers with long-distance underwater flowability applied to the horizontal tunnel parts of the trenches. (author)

  6. A proposal on restart rule of nuclear power plants with piping having local wall thinning subjected to an earthquake. Former part. Aiming at further application

    International Nuclear Information System (INIS)

    Urabe, Yoshio

    2011-01-01

    Restart rule of nuclear power plants (NPPs) with piping having local wall thinning subjected to an earthquake was proposed taking account of local wall thinning, seismic effects and restart of NPPs with applicability of 'Guidelines for NPP Response to an Earthquake (EPRI NP-6695)' in Japan. Japan Earthquake Damage Intensity Scale (JEDIS) and Earthquake Ground Motion Level (EGML) were introduced. JEDIS was classified into four scales obtained from damage level of components and structures of NPPs subjected to an earthquake, while EGML was divided into four levels by safe shutdown earthquake ground motion (So), elastic design earthquake ground motion (Sd) and design earthquake ground motion (Ss). Combination of JEDIS and EGML formulated 4 x 4 matrix and determined detailed conditions of restart of NPPs. As a response to an earthquake, operator walk inspections and evaluation of earthquake ground motion were conducted to know the level of JEDIS. JEDIS level requested respective allowable conditions of restart of NPP, which were scale level dependent and consisted of weighted combination of damage inspection (operator walk inspections, focused inspections/tests and expanded inspections), integrity evaluation and repair/replacement. If JEDIS were assigned greater than 3 with expanded inspections, inspection of piping with local wall thinning, its integrity evaluation and repair/replacement if necessary were requested. Inspection and evaluation of piping with local wall thinning was performed based on JSME or ASME codes. Detailed work flow charts were presented. Carbon steel piping and elbow was chosen for evaluation. (T. Tanaka)

  7. Study on Tensile Fatigue Behavior of Thermal Butt Fusion in Safety Class III High-Density Polyethylene Buried Piping in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Jong Sung; Lee, Young Ju; Oh, Young Jin

    2015-01-01

    High-density polyethylene (HDPE) piping, which has recently been applied to safety class III piping in nuclear power plants, can be butt-joined through the thermal fusion process, which heats two fused surfaces and then subject to axial pressure. The thermal fusion process generates bead shapes on the butt fusion. The stress concentrations caused by the bead shapes may reduce the fatigue lifetime. Thus, investigating the effect of the thermal butt fusion beads on fatigue behavior is necessary. This study examined the fatigue behavior of thermal butt fusion via a tensile fatigue test under stress-controlled conditions using finite element elastic stress analysis. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low-cycle fatigue region while having a negligible effect in the medium- and high-cycle fatigue regions

  8. Study on Tensile Fatigue Behavior of Thermal Butt Fusion in Safety Class III High-Density Polyethylene Buried Piping in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Sung; Lee, Young Ju [Sunchon National University, Suncheon (Korea, Republic of); Oh, Young Jin [KEPCO E and C, Yongin (Korea, Republic of)

    2015-01-15

    High-density polyethylene (HDPE) piping, which has recently been applied to safety class III piping in nuclear power plants, can be butt-joined through the thermal fusion process, which heats two fused surfaces and then subject to axial pressure. The thermal fusion process generates bead shapes on the butt fusion. The stress concentrations caused by the bead shapes may reduce the fatigue lifetime. Thus, investigating the effect of the thermal butt fusion beads on fatigue behavior is necessary. This study examined the fatigue behavior of thermal butt fusion via a tensile fatigue test under stress-controlled conditions using finite element elastic stress analysis. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low-cycle fatigue region while having a negligible effect in the medium- and high-cycle fatigue regions.

  9. 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

  10. Contributions of the ORNL piping program to nuclear piping design codes and standards

    International Nuclear Information System (INIS)

    Moore, S.E.

    1975-11-01

    The ORNL Piping Program was conceived and established to develop basic information on the structural behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design analysis and codes and standards. One of the objectives was to develop and qualify stress indices and flexibility factors for direct use in Code-prescribed design analysis methods. Progress in this area is described

  11. Development of nonlinear dynamic analysis program for nuclear piping systems

    International Nuclear Information System (INIS)

    Kamichika, Ryoichi; Izawa, Masahiro; Yamadera, Masao

    1980-01-01

    In the design for nuclear power piping, pipe-whip protection shall be considered in order to keep the function of safety related system even when postulated piping rupture occurs. This guideline was shown in U.S. Regulatory Guide 1.46 for the first time and has been applied in Japanese nuclear power plants. In order to analyze the dynamic behavior followed by pipe rupture, nonlinear analysis is required for the piping system including restraints which play the role of an energy absorber. REAPPS (Rupture Effective Analysis of Piping Systems) has been developed for this purpose. This program can be applied to general piping systems having branches etc. Pre- and post- processors are prepared in this program in order to easily input the data for the piping engineer and show the results optically by use of a graphic display respectively. The piping designer can easily solve many problems in his daily work by use of this program. This paper describes about the theoretical background and functions of this program and shows some examples. (author)

  12. Determination of leakage areas in nuclear piping

    International Nuclear Information System (INIS)

    Keim, E.

    1997-01-01

    For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack

  13. Determination of leakage areas in nuclear piping

    Energy Technology Data Exchange (ETDEWEB)

    Keim, E. [Siemens/KWU, Erlangen (Germany)

    1997-04-01

    For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack.

  14. Fatigue check of nuclear safety class 1 reactor coolant pipe

    International Nuclear Information System (INIS)

    Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

    2015-01-01

    Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

  15. Impact of inservice inspection on the reliability of nuclear piping

    International Nuclear Information System (INIS)

    Woo, H.H.

    1983-12-01

    The reliability of nuclear piping is a function of piping quality as fabricated, service loadings and environments, plus programs of continuing inspection during operation. This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident

  16. Computer aided design of piping for a radiochemical plant

    Energy Technology Data Exchange (ETDEWEB)

    Selvaraj, P G; Chandrasekhar, A; Chandrasekar, A V [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Raju, R P; Mahudeeswaran, K V; Kumar, S V [Reprocessing Group, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    In a radiochemical plant such as reprocessing plants, process equipment, storage tanks, liquid transfer systems and the associated pipe lines etc. are housed in series of concrete cells. Availability of limited cell space/volume, provision of various modes of liquid transfers with associated redundancies and instrumentation lines with standby alternatives increase the overall piping density. Designing such high density piping layout without interference is quite complex and needs lot of human efforts. This paper briefly describes development of computer codes for the entire scheme of design, drafting and fabrication of piping for nuclear fuel reprocessing plant. The general organisation of various programs, their functions, the complete sequence of the scheme and the flow of data are presented. High degree of reliability of each routine, considerable error checking facilities, marking legends on the drawings, provision for scaling in drafting and accuracy to the extent of one mm in layout design are some of the important features of this scheme. (author). 1 fig.

  17. Design specifications for ASME B and PV Code Section III nuclear class 1 piping

    International Nuclear Information System (INIS)

    Richardson, J.A.

    1978-01-01

    ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

  18. Fire protection in Angra-2 nuclear power plant. The use of fire protection collars on plastic piping systems

    International Nuclear Information System (INIS)

    Oliveira Segabinaze, R. de

    1994-01-01

    The object of this paper is to briefly the use of fire protection collars on plastic piping systems passing through wall and floor penetration. The fire protection collars consist of a stainless steel housing, in which the leading edges of two pivoting plates are in constant pressure contact with the pipe. In case of fire these plates react on the softened pipe with a guillotine action, thereby stopping the flow; within the housing a foam material expands to fill the space when subject to the heat of the fire. The piping project has to be modified to permit the fixing of the collars to walls and floor penetrations. (author). 2 refs, 9 figs

  19. Study on the Simulation of Crud Formation using Piping Materials of Nuclear Power Plant in High Temperature Water

    International Nuclear Information System (INIS)

    Kim, Sang Hyun; Kim, In Sup; Lee, Kun Jai

    2005-01-01

    High temperature - high pressure apparatus was developed to simulate nickel fewite corrosion products which were main compositions of the radioactive crud in the nuclear power plant. Corrosion product similar to the crud was obtained by a tube accumulator system. Nickel alloy (Inconel 690) and carbon steel (SA106 Gr. C) were corroded at 270 in the corrosion product generator. Ni ions and Fe ions dissolved by corrosion reaction were able to be transported to the accumulator because the crud generation mechanism was the solubility change with temperature. To evaluate the properties of simulated corrosion products, scanning electron microscope (SEM) observation and EDAX analysis were performed. SEM observation of corrosion product showed the needle like or crystal structure of oxide depending on precipitating location. The crystal oxide was the nickel ferrite, which was similar to the crud in nuclear power plants.

  20. Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-05-15

    The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

  1. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

    International Nuclear Information System (INIS)

    1984-08-01

    IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry

  2. Research and design of hanger and support series of nuclear safety class process piping

    International Nuclear Information System (INIS)

    Mao Chengzhang; Shi Jiemin

    1995-12-01

    Hangers and supports of nuclear safety class piping are an important part of primary system piping in a nuclear power plant. They will directly affect the reliability of operation, the period at construction and the investment for a nuclear power plant. It is an absolutely necessary job for Pakistan Chashma Nuclear Power Plant Project to research and design a series of piping supports in accordance with ASME-III NF. It is also an important designing for developing nuclear power plant later in China. After working over two years, a series of piping supports of nuclear safety class which have 57 types and more than 2460 specifications have been designed. This series is perfect, and can satisfy the requirements of piping final designing for nuclear power plant. This series of hangers and supports is mainly used in the process piping of nuclear safety class 1,2,3. They can also be used in other piping of nuclear safety class and piping with aseismic requirement of non-nuclear safety class

  3. The residual stress distribution in welded pipe inner surface of stainless steel from the nuclear power plant in Ringhals

    International Nuclear Information System (INIS)

    Larsson, L.E.

    1984-06-01

    The axial residual stress distribution on the inner surface of welded pipes of stainless steel SS 2333 (AISI 304) have been measured using the X-ray diffraction technique. Four halves of two pipes with the outer diameter of 114 mm and wall thickness of 10 mm were investigated. The result on the pipe inner surface shows compressive stresses in the weld metal and tensile stresses within a region between 8-23 mm with a maximum of 180MPa at a distance of 17 mm from the weld centerline. The maximum axial and circumferential residual stresses on the pipe outer surface are of the magnitude of 100 MPa. By cutting the pipes into two halves these stresses are relaxed by about 35 MPa. (author)

  4. Protection and isolation device for pipe maintenance, particularly for pipes of nuclear power plants. Dispositif d'isolement et de protection pour intervention sur tuyauterie, notamment tuyauterie de centrale nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Dohlen, G.; Le Marquis, J.C.; Oberlin, C.

    1984-09-28

    The device is aimed to be introduced and deployed inside a pipe, to collect and remove debris (dust or scraps) or foreign bodies, resulting from the work especially maintenance work being carried out. It comprises a central mast, a deformable sealing joint, mating with the interior of the conduit; a number of arms regularly distributed around the mast, which can be folded back against the mast, to permit introduction of the device into the conduit, each arm supporting the joint at one end and being pivoted on a common base at its other end; mechanical compression apparatus, connected to the mast and the base for deploying the apparatus, and flattening the joint against the interior surface of the conduit to which it is mated; and two sheets of material, each supported at its periphery by the joint, at least one of the sheets being suitable for isolating in sealed manner the space volumes which it delimits. The invention applies to maintenance operations for which the pipes have to be maintained under a controlled inert gas atmosphere, such as sodium circuits maintenance of nuclear power plants.

  5. Heat pipe nuclear reactor for space power

    Science.gov (United States)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  6. Nuclear piping system damping data studies

    International Nuclear Information System (INIS)

    Ware, A.G.; Arendts, J.G.

    1985-01-01

    A programm has been conducted at the Idaho National Engineering Laboratory to study structural damping data for nuclear piping systems and to evaluate if changes in allowable damping values for structural seismic analyses are justified. The existing pipe damping data base was examined, from which a conclusion was made that there were several sets of data to support higher allowable values. The parameters which most influence pipe damping were identified and an analytical investigation demonstrated that increased damping would reduce the required number of seismic supports. A series of tests on several laboratory piping systems was used to determine the effect of various parameters such as types of supports, amplitude of vibration, frequency, insulation, and pressure on damping. A multiple regression analysis was used to statistically assess the influence of the various parameters on damping, and an international pipe damping data bank has been formed. (orig.)

  7. Adaptation of the modern approaches for protection of nuclear power plants against the effects of postulated pipe ruptures to the Russian national guides. Problems and experience

    International Nuclear Information System (INIS)

    Berkovskij, A.; Kostarev, V.; Stevenson, J.D.

    2003-01-01

    Requirements for protection of Nuclear Power Plants against postulated ruptures of High-Energy Piping systems present practically in all National and International Guidelines for NPP Safety Design. Basically this problem consists of three general parts: (i) location of postulated ruptures; (2) consideration of the pipe rupture's consequences; and (3) realization of the protective measures. Presented paper describes the evolution and contemporary state of the problem regarding existing WWER NPPs in East Europe and Russia, as well as implementation of the High Energy Line Breaks (HELB) Analysis for the new-designed WWER Units. Paper presents the analysis of the current Russian National Guides regarding High Energy Line Breaks (HELB) problem. On the basis of this analysis the proposals for entering in Russian National Guide documentation changes and additions are developed. A special emphasis is given on the formulation of the intermediate rupture's locations based on the Strength Analysis according to PNAE G-7-002-86 (Russian Code) stress equations. The numerical comparative PNAE-ASME Analysis has been performed to illustrate the main approaches of the proposed methodology. (author)

  8. Nuclear power plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1976-01-01

    The invention concerns a quick-acting valve on the main-steam pipe of a nuclear power plant. The engineering design of the valve is to be improved. To the main valve disc, a piston-operated auxiliary valve disc is to be assigned closing a section of the area of the main valve disc. This way it is avoided that the drive of the main valve disc has to carry out different movements. 15 sub-claims. (UWI) [de

  9. Evaluation of thermal aging effect on primary pipe material in nuclear power plant by micro hardness test method

    International Nuclear Information System (INIS)

    Xue Fei; Yu Weiwei; Wang Zhaoxi; Ma Qinzheng; Liu Wei

    2012-01-01

    The investigation was carried out on the changes in mechanical properties of the primary pipe material Z3CN20.09M after 10000 h aging at 400℃ by using micro- Vickers and impact testing machine. The results show that the impact energy of testing material decreases. However, the micro-Vickers hardness of ferrite phase and austenite phase which constitute the testing material increase and keep constant, respectively. The intrinsic relations were analyzed between the micro-Vickers hardness and the impact energy to make an attempt to present the micro-Vickers hardness measurement as a method applicable to evaluating the thermal aging of the primary pipe material. (authors)

  10. A semi-analytical study of the vibrations induced by flow in the piping of nuclear power plants

    International Nuclear Information System (INIS)

    Maneschy, J.E.

    1981-01-01

    A semi-analytical method is presented to evaluate the piping system safety due to internal flow vibration excitation. The method is based on the application of a plane spectrum on the system, resulted by measured modal accelerations. A criteria is established to verify stress levels and compare with the allowable levels. (Author) [pt

  11. Numerical simulation and experimental verification of microstructure evolution in large forged pipe used for AP1000 nuclear power plants

    International Nuclear Information System (INIS)

    Wang, Shenglong; Yang, Bin; Zhang, Mingxian; Wu, Huanchun; Peng, Jintao; Gao, Yang

    2016-01-01

    Highlights: • Establish systematically the database of 316LN stainless steel for Deform-3D. • Simulate the microstructure evolution during forging of AP1000 primary coolant pipe. • Carry out full-scale forging experiment for verification in engineering practice. • Get desirable grain size in simulation and experiment. • The variation trends of grain sizes in simulation and experiment are consistent. - Abstract: AP1000 primary coolant pipe is a large special-shaped forged pipe made of 316LN stainless steel. Due to the non-uniform temperature and deformation during its forging, coarse and fine grains usually coexist in the forged pipe, resulting in the heterogeneous microstructure and anisotropic performance. To investigate the microstructure evolution during the entire forging process, in the present research, the database of the 316LN stainless steel was established and a numerical simulation was performed. The results indicate that the middle body section of the forged pipe has an extremely uniform average grain size with the value smaller than 30 μm. The grain sizes in the ends of body sections were ranged from 30 μm to 60 μm. Boss sections have relatively homogeneous microstructure with the average grain size 30 μm to 44 μm. Furthermore, a full-scale hot forging was carried out for verification. Comparison of theoretical and experimental results showed good agreement and hence demonstrated the capabilities of the numerical simulation presented here. It is noteworthy that all grains in the workpiece were confirmed less than 180 μm, which meets the designer’s demands.

  12. Survey on application of probabilistic fracture mechanics approach to nuclear piping

    International Nuclear Information System (INIS)

    Kashima, Koichi

    1987-01-01

    Probabilistic fracture mechanics (PFM) approach is newly developed as one of the tools to evaluate the structural integrity of nuclear components. This report describes the current status of PFM studies for pressure vessel and piping system in light water reactors and focuses on the investigations of the piping failure probability which have been undertaken by USNRC. USNRC reevaluates the double-ended guillotine break (DEGB) of rector coolant piping as a design basis event for nuclear power plant by using the PFM approach. For PWR piping systems designed by Westinghouse, two causes of pipe break are considered: pipe failure due to the crack growth and pipe failure indirectly caused by failure of component supports due to an earthquake. PFM approach shows that the probability of DEGB from either cause is very low and that the effect of earthquake on pipe failure can be neglected. (author)

  13. Aging and service wear of hydraulic and mechanical snubbers used on safety-related piping and components of nuclear power plants. Phase I study

    Energy Technology Data Exchange (ETDEWEB)

    Bush, S H; Heasler, P G; Dodge, R E

    1986-02-01

    This report presents an overview of hydraulic and mechanical snubbers used on nuclear piping systems and components, based on information from the literature and other sources. The functions and functional requirements of snubbers are discussed. The real versus perceived need for snubbers is reviewed, based primarily on studies conducted by a Pressure Vessel Research Committee. Tests conducted to qualify snubbers, to accept them on a case-by-case basis, and to establish their fitness for continued operation are reviewed. This report had two primary purposes. The first was to assess the effects of various aging mechanisms on snubber operation. The second was to determine the efficacy of existing tests in determining the effects of aging and degradation mechanisms. These tests include breakaway force, drag force, velocity/ acceleration range for activation in tension or compression, release rates within specified tension/compression limits, and restricted thermal movement. The snubber operating experience was reviewed using licensee event reports and other historical data for a period of more than 10 years. Data were statistically analyzed using arbitrary snubber populations. Value-impact was considered in terms of exposure to a radioactive environment for examination/ testing and the influence of lost snubber function and subsequent testing program expansion on the costs and operation of a nuclear power plant. The implications of the observed trends were assessed; recommendations include modifying or improving examination and testing procedures to enhance snubber reliability. Optimization of snubber populations by selective removal of unnecessary snubbers was also considered. (author)

  14. Utilizing clad piping to improve process plant piping integrity, reliability, and operations

    International Nuclear Information System (INIS)

    Chakravarti, B.

    1996-01-01

    During the past four years carbon steel piping clad with type 304L (UNS S30403) stainless steel has been used to solve the flow accelerated corrosion (FAC) problem in nuclear power plants with exceptional success. The product is designed to allow ''like for like'' replacement of damaged carbon steel components where the carbon steel remains the pressure boundary and type 304L (UNS S30403) stainless steel the corrosion allowance. More than 3000 feet of piping and 500 fittings in sizes from 6 to 36-in. NPS have been installed in the extraction steam and other lines of these power plants to improve reliability, eliminate inspection program, reduce O and M costs and provide operational benefits. This concept of utilizing clad piping in solving various corrosion problems in industrial and process plants by conservatively selecting a high alloy material as cladding can provide similar, significant benefits in controlling corrosion problems, minimizing maintenance cost, improving operation and reliability to control performance and risks in a highly cost effective manner. This paper will present various material combinations and applications that appear ideally suited for use of the clad piping components in process plants

  15. Development and testing of restraints for nuclear piping systems

    International Nuclear Information System (INIS)

    Kelly, J.M.; Skinner, M.S.

    1980-06-01

    As an alternative to current practice of pipe restraint within nuclear power plants it has been proposed to adopt restraints capable of dissipating energy in the piping system. The specific mode of energy dissipation focused upon in these studies is the plastic yielding of steels utilizing relative movement between the pipe and the base of the restraint, a general mechanism which has been proven as reliable in several allied studies. This report discusses the testing of examples of two energy-absorbing devices, the results of this testing and the conclusions drawn. This study concentrated on the specific relevant performance characteristics of hysteretic behavior and degradation with use. The testing consisted of repetitive continuous loadings well into the plastic ranges of the devices in a sinusoidal or random displacement controlled mode

  16. Effect of ferrite on the precipitation of σ phase in cast austenitic stainless steel used for primary coolant pipes of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yongqiang; Li, Na, E-mail: wangyongqiang1124@163.com [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology, Beijing (China)

    2017-11-15

    The effect of ferrite phase on the precipitation of σ phase in a Z3CN20.09M cast austenitic stainless steel (CASS) used for primary coolant pipes of pressurized water reactor (PWR) nuclear power plants was investigated by using isothermal heat-treatment, optical microscopy (OM), transmission electron microscopy (TEM) and electron probe microanalysis (EPMA) techniques. The influence of different morphologies and volume fractions of ferrite in the σ phase formation mechanism was discussed. The amount of σ phase precipitated in all specimens with different microstructures increased with increasing of aging time, however, the precipitation rate is significant different. The formation of σ phase in specimens with the coarsest ferrite and the dispersively smallest ferrite is slowest. The lowest level Cr content in ferrite and fewest α/γ interfaces in specimen are the main reasons for the slowest σ precipitation due to they are unfavorable for the kinetics and thermodynamics of phase transformation respectively. By contraries, the fastest formation of σ phase takes place in specimens with narrow and long ferrite due to the most α/γ interfaces and higher Cr content in ferrite which are beneficial for preferential nucleation and formation thermodynamics of σ phase. (author)

  17. Application of Nano-Structured Coatings for Mitigation of Flow-Accelerated Corrosion in Secondary Pipe Systems of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hyun; Kim, Jong Jin; Yoo, Seung Chang; Huh, Jae Hoon; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    Flow-accelerated corrosion (FAC) is a complex corrosion process combined with mechanical reaction with fluid. There were lots of research to mitigate FAC such as controlling temperature or water chemistry but in this research, we adopt active coating techniques especially nano-particle reinforced coatings. One of the general characteristics of FAC and its mitigation is that surface friction due to surface morphology makes a significant effect on FAC. Therefore to form a uniform coating layers, nano-particles including TiO2, SiC, Fe-Cr-W and Graphene were utilized. Those materials are known as greatly improve the corrosion resistance of substrates such as carbon steels but their effects on mitigation of FAC are not revealed clearly. Therefore in this research, the FAC resistive performance of nano-structured coatings were tested by electrochemical impedance spectroscopy (EIS) in room temperature 15 wt% sulfuric acid. As the flow-accelerated corrosion inhibitors in secondary piping system of nuclear power plants, various kinds of nano-structured coatings were prepared and tested in room-temperature electrochemical cells. SHS7740 with two types of Densifiers, electroless nickel plating with TiO2 are prepared. Electropolarization curves shows the outstanding corrosion mitigation performance of SHS7740 but EIS results shows the promising potential of Ni-P and Ni-P-TiO2 electroless nickel plating. For future work, high-temperature electrochemical analysis system will be constructed and in secondary water chemistry will be simulated.

  18. Application of nano-structured coatings to mitigate flow-accelerated corrosion in secondary pipe systems of nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Seung Hyun; Kim, Jong Jin; Yoo, Seung Chang; Kim, Ji Hyun

    2014-01-01

    Carbon steel is widely used as a structural material in secondary pipe systems. However, the passivity of carbon steel is not sufficient for protection in secondary water chemistry with a very fast-flowing fluid because of the dissolution of ferrous and magnetite ions and surface friction at the interface of the coolant and pipe surface. There have been many efforts to mitigate flow-accelerated corrosion through adoption of advanced water chemistries such as optimized dissolve oxygen (DO) concentration and temperature, as well as usage of new additives such as monoethanol amine (ETA) to adjust pH. However, these mitigation techniques pose certain challenges relating to the compatibility of new water chemistries with the steam generator, the thermal efficiency of the secondary side, etc. In this study, to improve the passivity of carbon steel, nanostructured coatings especially nanoparticle-enhanced surface coatings were adopted to improve resistance to corrosion and wear. Nanoparticles in the coating matrix help decrease the electrochemical potential compared coatings without nanoparticles, and thus help improve the mechanical properties, especially hardness, through precipitation. In other words, nanoparticle-enhanced surface coatings have the potential to mitigate flow-accelerated corrosion in secondary pipe systems. As candidate coatings, TiO 2 - and SiC-enhanced electrolytic and electroless nickel plating and Fe-Cr-W amorphous metallic coatings (AMC) were selected by acquiring the Pourbaix diagram with thermodynamic calculations. Both TiO 2 and SiC show a stable state in secondary water chemistry, and it is estimated that Fe-Cr-W can be applied to secondary water chemistry because it has a similar chemical composition to carbon steel. Electron microscopic analysis results with scanning electron microscopy (SEM) and tunneling electron microscopy (TEM) show the distribution of TiO 2 nanoparticles in the nickel matrix coating layer, whereas the SiC nanoparticles

  19. The development of three dimensional inspection and tracking system for the maintenance of pipes in the nuclear power plants

    International Nuclear Information System (INIS)

    Hwang, Suk Young; Kim, Chul Jung; Baik, Sung Hoon; Cho, Jai Wan; Park, Seung Kyu

    1999-12-01

    We developed 3D laser camera sensors for weld seam tracking and inspection of radioactive NPP pipes. The developed sensor's optical system adopts the optical triangulation method with the line beam generation and imaging optics. A laser line extraction algorithm accompanying preprocessing of noise reduction has been developed on images captured from the sensor. Experimental results validate the physical accuracy of the sensor hardware and the robustness of the image processing algorithms. A 3D shape reconstruction algorithm from multiple laser lines was proposed and the resulting 3D shape was visualized on the developed 3D graphic program environment utilizing OpenGL graphic libraries. And also, two D.O.F precise servo controlled mechanism was developed. The experimental results on weld seam tracking and inspection tasks show the practical feasibility of the developed sensors and the image processing algorithms. (author)

  20. The development of three dimensional inspection and tracking system for the maintenance of pipes in the nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Suk Young; Kim, Chul Jung; Baik, Sung Hoon; Cho, Jai Wan; Park, Seung Kyu

    1999-12-01

    We developed 3D laser camera sensors for weld seam tracking and inspection of radioactive NPP pipes. The developed sensor's optical system adopts the optical triangulation method with the line beam generation and imaging optics. A laser line extraction algorithm accompanying preprocessing of noise reduction has been developed on images captured from the sensor. Experimental results validate the physical accuracy of the sensor hardware and the robustness of the image processing algorithms. A 3D shape reconstruction algorithm from multiple laser lines was proposed and the resulting 3D shape was visualized on the developed 3D graphic program environment utilizing OpenGL graphic libraries. And also, two D.O.F precise servo controlled mechanism was developed. The experimental results on weld seam tracking and inspection tasks show the practical feasibility of the developed sensors and the image processing algorithms. (author)

  1. Nuclear-piping-repair planning today needs skill, organization

    International Nuclear Information System (INIS)

    O'Keefe, W.

    1986-01-01

    Nuclear power plant piping continues to experience failures and imminent threat of failure, despite a high level of care in design, analysis, fabrication, or installation. Continual inspection and surveillance and letter-by-letter following of procedures are not completely effective remedies, either. Both short-time-frame accidents and slowly progressing insidious complaints have caused loss of capacity, availability, and even confidence that the unit will work at close-to-expected performance. The fixes for nuclear-piping complaints cover a wide span, from mere carrying out of well-known repair procedures on either small scale or large, all the way to highly engineered solutions to a problem, with months of study and analysis followed by weighing of alternative methods. With some of the problems, little special planning is necessary. The repair is understood, and the time it needs is well within the envelope of a scheduled outage. Radiation exposure of personnel will not exceed expected moderate limits. And if the repair is a repeat performance of a recent similar one, little can go wrong. The planning for many other repairs, however, is so essential that even a minor failing in it will bring a debacle, with over-run, losses in revenue, and senseless expenditure of man-rems. Look at two types of planning for nuclear piping repair, as revealed at a recent American Welding Society conference on maintenance welding in nuclear power plants

  2. Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1983-06-01

    The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

  3. Structural integrity evaluation of nuclear piping cracket

    International Nuclear Information System (INIS)

    Cadiz Deleito, J.C.

    1985-01-01

    The methodology to evaluation of cracks in nuclear piping is exposed. Linear elastic fracture mechanic is used to prediction of growing crack and the net section collapse theory compared with acceptation criteria of both ASME III and ASME XI code. A case allowable under ASME XI criteria is analysed under ASME III requirements. Consideration must be given to local phenomenon in crack area and local stress evaluated and compared with ASME III acceptation criteria. (author)

  4. Corrosion Characteristics of Nano-structured Coatings for the Application in Secondary Piping System of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Won; Kim, Seung Hyun; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Coating surface using less corrosive metal is one of methods that reduce electrochemical corrosion. And metal oxide like a TiO{sub 2} is studied because it is stable, insoluble when coating is exposed severe environment. Several coating technics are used for better corrosion resistance. Pysical vapor deposition(PVD), chemical vapor deposition(CVD), thermal spray, electroplating, electroless etc. But thermal spray coating makes thermal stress to substrates because its temperature are more than 3000K. And powder's deformation can occur. And CVD makes decarburization near interface between surface and coating layer. In addition, CVD and PVD needs vacuum chamber. Electroplating is chemical reaction at surface, but it needs electric power. On the other hands, electroless plating dosen't needs electric power and it's temperature is low than thermal spray. Also the pipe dipping into the chemically solution can proceed coating easily. To reduce FAC, we have experiment about corrosion resistance of electroless Ni-P coated carbon steel in room temperature. And it has possibility of reducing corrosion and addition of TiO{sub 2} nano particles in Ni-P coating layer makes having better corrosion resistance. And results give us a possibility that electroless Ni-P coating added TiO{sub 2} nano particle can have better corrosion resistance compared carbon steel. So it needs study about high temperature corrosion experiment of electroless Ni-P coating added TiO{sub 2} nano particle.

  5. Application of heat pipes in nuclear reactors for passive heat removal

    Energy Technology Data Exchange (ETDEWEB)

    Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

  6. Summary and accomplishments of the ORNL program for nuclear piping design criteria

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1975-11-01

    The ORNL Piping Program was defined and established to develop basic information on the structure behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design codes and standards. Charts are presented showing the percentage completion of the various program tasks

  7. 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

  8. Application of bounding spectra to seismic design of piping based on the performance of above ground piping in power plants subjected to strong motion earthquakes

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-02-01

    This report extends the potential application of Bounding Spectra evaluation procedures, developed as part of the A-46 Unresolved Safety Issue applicable to seismic verification of in-situ electrical and mechanical equipment, to in-situ safety related piping in nuclear power plants. The report presents a summary of earthquake experience data which define the behavior of typical U.S. power plant piping subject to strong motion earthquakes. The report defines those piping system caveats which would assure the seismic adequacy of the piping systems which meet those caveats and whose seismic demand are within the bounding spectra input. Based on the observed behavior of piping in strong motion earthquakes, the report describes the capabilities of the piping system to carry seismic loads as a function of the type of connection (i.e. threaded versus welded). This report also discusses in some detail the basic causes and mechanisms for earthquake damages and failures to power plant piping systems

  9. 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

  10. Buried piping integrity management at fossil power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shulder, Stephen J. [Structural Integrity Associates, Annapolis, MD (United States); Biagiotti, Steve [Structural Integrity Associates, Inc., Centennial, CO (United States)

    2011-07-15

    In the last decade several industries (oil and gas pipelines, nuclear power, and municipal water) have experienced an increase in the frequency and public scrutiny of leaks and failures associated with buried piping and tank assets. In several industries, regulatory pressure has resulted in the mandated need for databases and inspection programs to document and ensure the continued integrity of these assets. Power plants are being extended beyond their design life and the condition of below grade assets is essential toward continued operation. This article shares the latest advances in managing design, operation, process, inspection, and historical data for power plant piping. Applications have also been developed to help with risk prioritization, inspection method selection, managing cathodic protection data for external corrosion control, and a wide variety of other information. This data can be managed in a GIS environment allowing two and three dimensional (2D and 3D) access to the database information. (orig.)

  11. Operating Experience Insights into Pipe Failures for Electro-Hydraulic Control and Instrument Air Systems in Nuclear Power Plant. A Topical Report from the Component Operational Experience, Degradation and Ageing Programme

    International Nuclear Information System (INIS)

    2015-01-01

    Structural integrity of piping systems is important for plant safety and operability. In recognition of this, information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organisations (e.g. OECD/NEA and IAEA) and industry organisations worldwide to provide systematic feedback for example to reactor regulation and research and development programmes associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programmes, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. Several OECD member countries have agreed to establish the OECD/NEA 'Component Operational Experience, Degradation and Ageing Programme' (CODAP) to encourage multilateral co-operation in the collection and analysis of data relating to degradation and failure of metallic piping and non-piping metallic passive components in commercial nuclear power plants. The scope of the data collection includes service-induced wall thinning, part through-wall cracks, through-wall cracks with and without active leakage, and instances of significant degradation of metallic passive components, including piping pressure boundary integrity. The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) acts as an umbrella committee of the Project. CODAP is the continuation of the 2002-2011 'OECD/NEA Pipe Failure Data Exchange Project' (OPDE) and the Stress Corrosion Cracking Working Group of the 2006-2010 'OECD/NEA Stress Corrosion Cracking and Cable Ageing Project' (SCAP). OPDE was formally launched in May 2002. Upon completion of the third term (May 2011), the OPDE project was officially closed to be succeeded by CODAP. SCAP was enabled by a voluntary contribution from Japan. It was formally launched in June 2006 and officially closed with an international workshop held in Tokyo in May

  12. Evaluation of seismic margins for an in-plant piping system

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

  13. Program to justify life extension of older nuclear piping systems

    International Nuclear Information System (INIS)

    Burr, T.K.; Dwight, J.E. Jr.; Morton, D.K.

    1991-01-01

    The Idaho National Engineering Laboratory (INEL) has a history of more than 40 years devoted to the operation of nuclear reactors designed for research and experiments. The Advanced Test Reactor (ATR) is one such operating reactor whose mission requires continued operation for an additional 25 years or more. Since the ATR is approaching its design life of twenty years, life extension evaluations have been initiated. Of particular importance are the associated high temperature, high pressure loop piping system supporting in--reactor experiments. Failure of this piping could challenge core safety margins. Since regulatory rules for nuclear power plant life extension are only in the formulation stage, the current technical guidance on this subject provided by the Department of Energy (DOE) or the commercial nuclear industry is incomplete. In the interim, order to assure continued safe operation of this piping beyond its initial design life, a program has been developed to provide the necessary technical justification for life extension. This paper describes a program that establishes Section 11 of the ASME Boiler and Pressure Vessel Code as the governing criteria document, retains B31.1 as the Code of record for Section 11 activities, specifies additional inservice inspection requirements more strict than Section 11, and relies heavily on flaw detection and fracture mechanics evaluations. 18 refs., 2 figs

  14. Basic concepts about application of dual vibration absorbers to seismic design of nuclear piping systems

    International Nuclear Information System (INIS)

    Hara, F.; Seto, K.

    1987-01-01

    The design value of damping for nuclear piping systems is a vital parameter in ensuring safety in nuclear plants during large earthquakes. Many experiments and on-site tests have been undertaken in nuclear-industry developed countries to determine rational design values. However damping value in nuclear piping systems is so strongly influenced by many piping parameters that it shows a tremendous dispersion in its experimental values. A new trend has recently appeared in designing nuclear pipings, where they attempt to use a device to absorb vibration energy induced by seismic excitation. A typical device is an energy absorbing device, made of a special material having a high capacity of plasticity, which is installed between the piping and the support. This paper deals with the basic study of application of dual vibration absorbers to nuclear piping systems to accomplish high damping value and reduce consequently seismic response at resonance frequencies of a piping system, showing their effectiveness from not only numerical calculation but also experimental evaluation of the vibration responses in a 3D model piping system equipped with dual two vibration absorbers

  15. Safety evaluation of socket weld integrity in nuclear piping

    International Nuclear Information System (INIS)

    Choi, Y.H.; Kim, H.J.; Choi, S.Y.; Kim, Y.J.; Kim, Y.J.

    2004-01-01

    The purposes of this paper are to evaluate the integrity of socket weld in nuclear piping and prepare the technical basis for a new guideline on radiographic testing (RT) for the socket weld. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because lots of failures and leaks have been reported in the socket weld. The root causes of the socket weld failure are known as unanticipated loadings such as vibration or thermal fatigue and improper weld joint during construction. The ASME Code sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld. Many failure cases, however, showed that the gap requirement was not satisfied. The Code also requires magnetic particle examination (MT) or liquid penetration examination (PT) on the socket weld, but not radiographic examination (RT). It means that it is not easy to examine the 1/16 inch gap in the socket weld by using the NDE methods currently required in the Code. In this paper, the effects of the requirements in the ASME Code sec. III on the socket weld integrity were evaluated by using finite element method. The crack behavior in the socket weld was also investigated under vibration event in nuclear power plants. The results showed that the socket weld was very susceptible to the vibration if the requirements in ASME Code were not satisfied. The constraint between the pipe and fitting due to the contact significantly affects the integrity of the socket weld. This paper also suggests a new guideline on the RT for the socket weld during construction stage in nuclear power plants. (orig.)

  16. 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

  17. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 5. Summary - Piping Review Committee conclusions and recommendations

    International Nuclear Information System (INIS)

    1985-04-01

    This document summarizes a comprehensive review of NRC requirements for Nuclear Piping by the US NRC Piping Review Committee. Four topical areas, addressed in greater detail in Volumes 1 through 4 of this report, are included: (1) Stress Corrosion Cracking in Piping of Boiling Water Reactor Plants; (2) Evaluation of Seismic Design; (3) Evaluation of Potential for Pipe Breaks; and (4) Evaluation of Other Dynamic Loads and Load Combinations. This volume summarizes the major issues, reviews the interfaces, and presents the Committee's conclusions and recommendations for updating NRC requirements on these issues. This report also suggests research or other work that may be required to respond to issues not amenable to resolution at this time

  18. Through wall degradation problem of the turbine extraction steam drain piping due to liquid drop impingement and measures taken for this problem at Fukushima Dai-ichi Nuclear Power Plant Unit 6

    International Nuclear Information System (INIS)

    Inagaki, Takeyuki; Kobayashi, Teruaki; Shimada, Shigeru; Inoue, Ryousuke; Usuba, Satoshi; Kimura, Takeo

    2011-01-01

    Through wall degradation was found on the extraction steam drain piping of Unit 6 of Fukushima Dai-ichi Nuclear Power Plant owned by Tokyo Electric Power Company after replacement of the turbine rotors with those of higher thermal efficiency. The mechanism of this degradation was loss of material due to liquid drop impingement. Since the estimated life time of the piping based on wall thickness measurements before the replacement was at least 9 years, the rapid wall thinning occurred after the replacement. This paper describes a summary of the phenomenon, its degradation mechanism and root cause, a temporary measurement taken for an immediate action and permanent measures taken during the next refueling outage. (author)

  19. Thermal expansion movements of piping during FFTF plant startup

    International Nuclear Information System (INIS)

    Lindquist, M.R.

    1981-03-01

    FFTF liquid metal piping exhibits significant displacements during heatup of the plant heat transport system. Verification of correct piping movements is important to assure that no restraints are present and to provide data for additional piping design/analysis validation. A test program is described in which a series of measurements were taken at selected piping locations. These data were obtained during Plant Acceptance Testing involving system heatup cycles to approximately 800 0 F(427 0 C). Typical test data are shown and compared to analytical predictions. Two piping system problems that were identified as a result of the testing are described along with resolutions thereof. Establishment of final baseline data is discussed

  20. Comparison of piping models for digital power plant simulators

    International Nuclear Information System (INIS)

    Sowers, G.W.

    1979-08-01

    Two piping models intended for use in a digital power plant simulator are compared. One is a finite difference approximation to the partial differential equation called PIPE, and the other is a function subroutine that acts as a delay operator called PDELAY. The two models are compared with respect to accuracy and execution time. In addition, the stability of the PIPE model is determined. The PDELAY model is found to execute faster than the PIPE model with comparable accuracy

  1. Criteria for accepting piping vibrations measured during FFTF plant startup

    International Nuclear Information System (INIS)

    Huang, S.N.

    1981-03-01

    Piping in the Fast Flux Test Facility is subjected to low-amplitude, high cycle vibration over the plant lifetime. Excitation sources include the mechanical vibration induced by main centrifugal pumps, auxiliary reciprocating pumps, EM pumps and possible flow oscillations. Vibration acceptance criteria must be established which will prevent excessive pipe and support fatigue damage when satified. This paper describes the preparation of such criteria against pipe failure used for acceptance testing of the Fast Flux Test Facility main heat transport piping

  2. Leak-before-break behaviour of nuclear piping systems

    International Nuclear Information System (INIS)

    Bartholome, G.; Wellein, R.

    1992-01-01

    The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs

  3. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  4. Nuclear power plants

    International Nuclear Information System (INIS)

    Kiyokawa, Teruyuki; Soman, Yoshindo.

    1985-01-01

    Purpose: To constitute a heat exchanger as one unit by integrating primary and secondary coolant circuits with secondary coolant circuit and steam circuit into a single primary circuit and steam circuit. Constitution: A nuclear power plant comprises a nuclear reactor vessel, primary coolant pipeways and a leakage detection system, in which a dual-pipe type heat exchanger is connected to the primary circuit pipeway. The heat conduction tube of the heat exchanger has a dual pipe structure, in which the inside of the inner tube is connected to the primary circuit pipeway, the outside of the outer tube is connected to steam circuit pipeway and a fluid channel is disposed between the inner and outer tubes and the fluid channel is connected to the inside of an expansion tank for intermediate heat medium. The leak detection system is disposed to the intermediate heat medium expansion tank. Sodium as the intermediate heat medium is introduced from the intermediate portion (between the inner and outer tubes) by way of inermediate heat medium pipeways to the intermediate heat medium expansion tank and, further, to the intermediate portion for recycling. (Kawakami, Y.)

  5. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1993-09-01

    Quarterly reports on the operation of Finnish nuclear power plants describe events and observations, relating to nuclear safety and radiation protection which the Finnish Centre for Radiation and Nuclear Safety considers safety significant. Safety-enhancing modifications at the nuclear power plants and issues relating to the use of nuclear energy which are of general interest are also reported. The reports include a summary of the radiation safety of plant personnel and the environment, as well as tabulated data on the production and load factors of the plants. In the first quarter of 1993, a primary feedwater system pipe break occurred at Loviisa 2, in a section of piping after a feedwater pump. The break was erosion-corrosion induced. Repairs and inspections interrupted power generation for seven days. On the International Nuclear Event Scale the event is classified as a level 2 incident. Other events in the first quarter of 1993 had no bearing on nuclear safety and radiation protection

  6. B Plant process piping replacement feasibility study

    International Nuclear Information System (INIS)

    Howden, G.F.

    1996-01-01

    Reports on the feasibility of replacing existing embedded process piping with new more corrosion resistant piping between cells and between cells and a hot pipe trench of a Hanford Site style canyon facility. Provides concepts for replacement piping installation, and use of robotics to replace the use of the canyon crane as the primary means of performing/supporting facility modifications (eg, cell lining, pipe replacement, equipment reinstallation) and operational maintenenace

  7. 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

  8. Bolted Flanged Connection for Critical Plant/Piping Systems

    International Nuclear Information System (INIS)

    Efremov, Anatoly

    2006-01-01

    A novel type of Bolted Flanged Connection with bolts and gasket manufactured on a basis of advanced Shape Memory Alloys is examined. Presented approach combined with inverse flexion flange design of plant/piping joint reveals a significant increase of internal pressure under conditions of a variety of operating temperatures relating to critical plant/piping systems. (author)

  9. Nuclear power plant information management system 'NUPIMAS'

    International Nuclear Information System (INIS)

    Matsumoto, M.; Saruyama, I.; Kurokawa, Y.; Kayano, M.; Katto, S.

    1980-01-01

    NUPIMAS is an interactive computer graphic system used for the design of pipings of nuclear power plant and the production of their drawings. Data on piping, duct, cable tray, equipment and building are stored in the computer and the following conversational-mode design works are performed online by means of graphic display, plotter and others: (1) Piping route study and interference check. (2) Modification of piping route and specifications. (3) Semi-automatic design of low-temperature piping supports. As the result of these design works the following drawings and lists are produced and interactively refined by computer: (1) Composite drawings. (2) Piping assembly drawings and shop drawings. (3) Bill of material. (4) Welding procedure instruction. (5) Duct route drawings (Isometric and 3-plane views). (6) Shop and assembly drawings of supports, etc. This system is already in practical use, obtaining good results. (author)

  10. Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

    International Nuclear Information System (INIS)

    Lin, Y.C.; Xie, Y.J.; Wang, X.H.

    2004-01-01

    Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software

  11. Socket weld integrity in nuclear piping under fatigue loading condition

    International Nuclear Information System (INIS)

    Choi, Young Hwan; Choi, Sun Yeong

    2007-01-01

    The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading. The integrity of socket weld is regarded as a safety concern in nuclear power plants because many failures have been world-widely reported in the socket weld. Recently, socket weld failures in the chemical and volume control system (CVCS) and the primary sampling system (PSS) were reported in Korean nuclear power plants. The root causes of the socket weld failures were known as the fatigue due to the pressure and/or temperature loading transients and the vibration during the plant operation. The ASME boiler and pressure vessel (B and PV) Code Sec. III requires 1/16 in. gap between the pipe and fitting in the socket weld with the weld leg size of 1.09 x t 1 , where t 1 is the pipe wall thickness. Many failure cases, however, showed that the gap requirement was not satisfied. In addition, industry has demanded the reduction of weld leg size from 1.09 x t 1 to 0.75 x t 1 . In this paper, the socket weld integrity under the fatigue loading was evaluated using three-dimensional finite element analysis considering the requirements in the ASME Code. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P = 0 to 15.51 MPa, and the thermal transient ranging from T = 25 to 288 deg. C were considered. The results are as follows; (1) the socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME operation and maintenance (OM) code. (2) The effect of pressure or temperature transient load on socket weld in CVCS and PSS is not significant owing to the low frequency of transient during plant operation. (3) 'No gap' is very risky to the socket weld integrity for the systems having the vibration condition to exceed the requirement specified in the ASME OM Code and/or the transient loading condition from P = 0 and T = 25 deg. C to P = 15.51 MPa and T = 288 deg. C. (4

  12. Structural and stress analysis of nuclear piping systems

    International Nuclear Information System (INIS)

    Hata, Hiromichi

    1982-01-01

    The design of the strength of piping system is important in plant design, and its outline on the example of PWRs is reported. The standards and guides concerning the design of the strength of piping system are shown. The design condition for the strength of piping system is determined by considering the requirements in the normal operation of plants and for the safety design of plants, and the loads in normal operation, testing, credible accident and natural environment are explained. The methods of analysis for piping system are related to the transient phenomena of fluid, piping structure and local heat conduction, and linear static analysis, linear time response analysis, nonlinear time response analysis, thermal stress analysis and fluid transient phenomenon analysis are carried out. In the aseismatic design of piping system, it is desirable to avoid the vibration together with a building supporting it, and as a rule, to make it into rigid structure. The piping system is classified into high temperature and low temperature pipings. The formulas for calculating stress and the allowable condition, the points to which attention must be paid in the design of piping strength and the matters to be investigated hereafter are described. (Kako, I.)

  13. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  14. Titanium Loop Heat Pipes for Space Nuclear Radiators, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project will develop titanium Loop Heat Pipes (LHPs) that can be used in low-mass space nuclear radiators, such as...

  15. Survey of heat-pipe application under nuclear environment

    International Nuclear Information System (INIS)

    Tsuyuzaki, Noriyoshi; Saito, Takashi; Okamoto, Yoshizo; Hishida, Makoto; Negishi, Kanji.

    1986-11-01

    Heat pipes today are employed in a wide variety of special heat transfer applications including nuclear reactor. In this nuclear technology area in Japan, A headway speed of the heat pipe application technique is not so high because of safety confirmation and investigation under each developing step. Especially, the outline of space craft is a tendency to increase the size. Therefore, the power supply is also tendency to increase the outlet power and keep the long life. Under SP-100 project, the development of nuclear power supply system which power is 1400 - 1600 KW thermal and 100 KW electric power is steadily in progress. Many heat pipes are adopted for thermionic conversion and coolant system in order to construct more safety and light weight system for the project. This paper describes the survey of the heat pipe applications under the present and future condition for nuclear environment. (author)

  16. Lead plant application of leak-before-break to high energy piping. Final report, January 1989

    International Nuclear Information System (INIS)

    1989-01-01

    This report presents the experience gained during a successful application of a leak-before-break program by Duquesne Light Company. This program was directed at the high energy nuclear piping at Beaver Valley Power Station - Unit 2. This experience can be applied to other nuclear plant leak-before-break efforts in order to minimize the number of pipe whip restraints, jet impingement shields, snubbers, and to discount the consideration of remaining pipe rupture dynamic effects. The chronology of events leading to Nuclear Regulatory Commission approval of the Beaver Valley Power Station - Unit 2 lead plant effort is described. The final report and pertinent sections of the final Safety Evaluation Report are also included. (author)

  17. Nuclear piping criteria for Advanced Light-Water Reactors, Volume 1--Failure mechanisms and corrective actions

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This WRC Bulletin concentrates on the major failure mechanisms observed in nuclear power plant piping during the past three decades and on corrective actions taken to minimize or eliminate such failures. These corrective actions are applicable to both replacement piping and the next generation of light-water reactors. This WRC Bulletin was written with the objective of meeting a need for piping criteria in Advanced Light-Water Reactors, but there is application well beyond the LWR industry. This Volume, in particular, is equally applicable to current nuclear power plants, fossil-fueled power plants, and chemical plants including petrochemical. Implementation of the recommendations for mitigation of specific problems should minimize severe failures or cracking and provide substantial economic benefit. This volume uses a case history approach to high-light various failure mechanisms and the corrective actions used to resolve such failures. Particular attention is given to those mechanisms leading to severe piping failures, where severe denotes complete severance, large ''fishmouth'' failures, or long throughwall cracks releasing a minimum of 50 gpm. The major failure mechanisms causing severe failure are erosion-corrosion and vibrational fatigue. Stress corrosion cracking also has been a common problem in nuclear piping systems. In addition thermal fatigue due to mixing-tee and to thermal stratification also is discussed as is microbiologically-induced corrosion. Finally, water hammer, which represents the ultimate in internally-generated dynamic high-energy loads, is discussed

  18. 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.

  19. 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

  20. Nuclear power plant pressure vessels. Inservice inspections

    International Nuclear Information System (INIS)

    1995-01-01

    The requirements for the planning and reporting of inservice inspections of nuclear power plant pressure vessels are presented. The guide specifically applies to inservice inspections of Safety class 1 and 2 nuclear power plant pressure vessels, piping, pumps and valves plus their supports and reactor pressure vessel internals by non- destructive examination methods (NDE). Inservice inspections according to the Pressure Vessel Degree (549/73) are discussed separately in the guide YVL 3.0. (4 refs.)

  1. Seismic stops for nuclear power plants

    International Nuclear Information System (INIS)

    Cloud, R.L.; Leung, J.S.M.; Anderson, P.H.

    1989-01-01

    In the regulated world of nuclear power, the need to have analytical proof of performance in hypothetical design-basis events such as earth quakes has placed a premium on design configurations that are mathematically tractable and easily analyzed. This is particularly true for the piping design. Depending on how the piping analyses are organized and on how old the plant is, there may be from 200 to 1000 separate piping runs to be designed, analyzed, and qualified. In this situation, the development of snubbers seemed like the answer to a piping engineer's prayer. At any place where seismic support was required but thermal motion had to be accommodated, a snubber could be specified. But, as experience has now shown, the program was solved only on paper. This article presents an alternative to conventional snubbers. These new devices, termed Seismic Stops are designed to replace snubbers directly and look like snubbers on the outside. But their design is based on a completely different principle. The original concept has adapted from early seismic-resistant pipe support designs used on fossil power plants in California. The fundamental idea is to provide a space envelope in which the pipe can expand freely between the hot and cold positions, but cannot move outside the envelope. Seismic Stops are designed to transmit any possible impact load, as would occur in an earthquake, away from the pipe itself to the Seismic Stop. The Seismic Stop pipe support is shown

  2. A cost summary applicable to seismic construction and maintenance of nuclear safety related piping

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-01-01

    This paper presents a summary of costs applicable to nuclear power plant piping for an earthquake defined as 0.2 SSE-PGA as a function of three eras of initial construction: 1967--1974, 1974--1981 and 1981--1990. Costs have been presented for both new construction and maintenance in operating plants using both the original PSAR-FSAR design criteria and current SRP requirements. It is recommended that the cost information contained in this report be considered in evaluating the cost benefit relationships associated with current and proposed future changes in seismic design procedures applicable to safety-related piping systems

  3. State-of-practice review of ultrasonic in-service inspection of Class I system piping in commercial nuclear power plants

    International Nuclear Information System (INIS)

    Morris, C.J.; Becker, F.L.

    1982-08-01

    The Pacific Northwest Laboratory conducted a survey to determine the state of practice of ultrasonic in-service inspection of primary system piping in light water reactors. Personnel at four utilities, five inspection organizations, and three domestic reactor manufacturers were interviewed. The intention of the study was to provide a better understanding of the actual practices employed in in-service inspection of primary system piping and of the difficulties encountered

  4. Nuclear power plants

    International Nuclear Information System (INIS)

    Ushijima, Susumu.

    1984-01-01

    Purpose: To enable to prevent the degradation in the quality of condensated water in a case where sea water leakage should occur in a steam condenser of a BWR type nuclear power plant. Constitution: Increase in the ion concentration in condensated water is detected by an ion concentration detector and the leaking factor of sea water is calculated in a leaking factor calculator. If the sea water leaking factor exceeds a predetermined value, a leak generation signal is sent from a judging device to a reactor power control device to reduce the reactor power. At ehe same tiem, the leak generation signal is also sent to a steam condenser selection and isolation device to interrupt the sea water pump of a specified steam condenser based on the signal from the ion concentration detector, as well as close the inlet and outlet valves while open vent and drain valves to thereby forcively discharge the sea water in the cooling water pipes. This can keep the condensate desalting device from ion breaking and prevent the degradation in the quality of the reactor water. (Horiuchi, T.)

  5. Recent experience in nuclear plant nondestructive examinations

    International Nuclear Information System (INIS)

    Epps, T.N.

    1986-01-01

    This paper reviews recent experience in nuclear plant inservice inspection activities including ultrasonic examination of piping materials, personnel qualification, results, and the overall significance to the industry. Several areas of concern to the nuclear power industry have recently been addressed by Southern Company Services' (SCS) Inspection, Testing, and Engineering Department during implementation of preservice and in-service inspection activities in the SCS system. The most significant of these activities is the ultrasonic inspection of Type 304 stainless steel piping for the presence of intergranular stress corrosion cracking (IGSCC). This activity has been in the forefront of boiling water reactor (BWR) in-service inspections for the past several years

  6. Managing the Cost of Plant Piping System Leakage

    International Nuclear Information System (INIS)

    Jenco, John M.; Keck, Donna R.; Johnson, Gary L.

    2002-01-01

    Recent studies have shown that the average annual cost impact of external piping system leakage on commercial nuclear plant operations and maintenance can easily range into the millions of dollars for each reactor unit. Evidence suggests that this significant O and M cost reduction opportunity has largely been overlooked, due to the number of diverse line items and budget areas affected. Results released last year from an EPRI pilot study of more than a dozen reactor units at seven plant sites operated by multiple utilities found that the average annual cost impact was indeed around $1.6 million per year per unit. Subsequent field experience has also demonstrated that an effective fluid leak management program can substantially reduce these costs within the first three years of implementation. This paper presents the general cost impact research results from various studies, outlines key elements of an effective plant fluid leak management program, discusses important implementation issues, and presents results from case studies covering different utility approaches to developing and implementing an effective fluid leak management program. Actual cost data will be included where appropriate. (authors)

  7. Evaluation of biocide efficacy on microbiological induced corrosion of pipes and equipment from the 'process water system' of Embalse nuclear power plant (CNE)

    International Nuclear Information System (INIS)

    Forte Giacobone, A F; Burkart, A L; Pizarro, R; Rodriguez S; Belloni, M; Croatto, F; Ferrari, F; Herrera, C; Mendizabal, M; Montes, J; Rodriguez Aliciardi, M; Saucedo, R; Ovando, L

    2012-01-01

    In order to improve water quality, and mitigate recurrent bio corrosion phenomena affecting the components of the Process Water System of the CNE, a combined water treatment adding a commercial biocide product, based on bromide, to the currently injected chlorine was proposed. The aim of this study was to evaluate the effect of the added biocide on the kinetics of biofilm formation and growth, which is the precursor process to microbiological corrosion, and on the corrosion rates of carbon steel of pipes, heat exchanger shells and other system devices. For this purpose, a test bench was designed and built, reproducing the flow conditions at certain parts of the system. This facility was installed in the filtration shed of the Water Plant of the CNE. The test bench consisted of two parallel chambers, I and II, each in turn divided into a section for determining biofilm growth and corrosion rates of carbon steel coupons and another one to measure the kinetics of biofilm growth on stainless steel coupons. Both chambers received lake water chlorinated for 15 minutes each day. The chamber II received also the biocide. The corrosion rate in carbon steel coupons was evaluated by weight loss and Linear Polarization Resistance (LPR) measurements. The kinetics of biofilm growth on carbon steel coupons was measured using disruptive methods followed by quantification of the protein and carbohydrate content as an estimation of total biomase. The following bacterial groups were quantified through the dilution-extinction method: total aerobic bacteria, acid-producing bacteria, total anaerobic bacteria, sulfate reducing bacteria and bacteria precipitating iron and manganese. On the stainless steel coupons, the percent of coverage was evaluated by epi fluorescence microscopy. The corrosion rate results obtained both by weight loss as by LPR, showed no significant differences between both chambers, with and without biocide. Regarding the kinetics of biofilm growth on carbon steel

  8. Identification and reduction of piping-vibrations in plants

    Energy Technology Data Exchange (ETDEWEB)

    Kerkhof, K. [Stuttgart Univ. (Germany). MPA

    2012-07-01

    Safe operation, availability and lifetime assessment of piping systems are of utmost concern for plant operators. The use of tuned mass dampers is a rather new approach for reducing vibrations to avoid high cycle fatigue in a large chemical piping system. The investigated piping system is supported by a tall structure fixed at the base. As a result, the steel building stiffness decreases with height. Furthermore large piping-elbow forces act at the top of the building, which lead to large vibration amplitudes. Since both piping system and supporting structure exhibited these large vibration amplitudes, dampers or shock absorbers placed between them would prove ineffective. Therefore, special vibration absorbers were developed for such piping systems. The paper presents the design process, starting with an extensive system investigation up to the passive multi-axial vibration absorber design parameters. This includes: Laboratory tests with a mock-up pipe system, where the first design ideas for new passive vibration absorbers were investigated. Vibration measurements were carried out to investigate the current state of the vibration behaviour. The piping system was inspected; strain gauges were used to identify stress concentrations at welds and other notches due to ovalization. Finite element calculations were performed, first as a combined beam and shell model for the pipe without the support structure. A detailed model for the combined steel construction and pipe system was created. Model-updating was done to fit the calculated model to the experimental modal analysis data. Loading assumptions describing excitation forces from the mass flow were checked. Harmonic frequency analysis was performed. On the basis of these calculations design parameters for the passive vibration absorber were determined. Finally, a solution for the design of two passive vibration absorbers will be presented.

  9. Identification and reduction of piping-vibrations in plants

    International Nuclear Information System (INIS)

    Kerkhof, K.

    2012-01-01

    Safe operation, availability and lifetime assessment of piping systems are of utmost concern for plant operators. The use of tuned mass dampers is a rather new approach for reducing vibrations to avoid high cycle fatigue in a large chemical piping system. The investigated piping system is supported by a tall structure fixed at the base. As a result, the steel building stiffness decreases with height. Furthermore large piping-elbow forces act at the top of the building, which lead to large vibration amplitudes. Since both piping system and supporting structure exhibited these large vibration amplitudes, dampers or shock absorbers placed between them would prove ineffective. Therefore, special vibration absorbers were developed for such piping systems. The paper presents the design process, starting with an extensive system investigation up to the passive multi-axial vibration absorber design parameters. This includes: Laboratory tests with a mock-up pipe system, where the first design ideas for new passive vibration absorbers were investigated. Vibration measurements were carried out to investigate the current state of the vibration behaviour. The piping system was inspected; strain gauges were used to identify stress concentrations at welds and other notches due to ovalization. Finite element calculations were performed, first as a combined beam and shell model for the pipe without the support structure. A detailed model for the combined steel construction and pipe system was created. Model-updating was done to fit the calculated model to the experimental modal analysis data. Loading assumptions describing excitation forces from the mass flow were checked. Harmonic frequency analysis was performed. On the basis of these calculations design parameters for the passive vibration absorber were determined. Finally, a solution for the design of two passive vibration absorbers will be presented.

  10. The development of design method of nuclear piping system supported by elasto-plastic support structures (part 2)

    International Nuclear Information System (INIS)

    Endo, R.; Murota, M.; Kawabata, J-I.; Hirose, J.; Nekomoto, Y.; Takayama, Y.; Kobayashi, H.

    1995-01-01

    The conventional seismic design method of nuclear piping system is very conservative because of the accumulation of various safety factors in the design process, and nuclear piping systems are thought to have a large safety margin. Considering this situations, research program was promoted to furthermore rationalize nuclear power plants by reducing the amount of support structures and reducing the piping's seismic response through vibration energy absorption resulting from the elasto-plastic behavior of piping support structures. The research had the following three stages. In the first stage, we selected conventional piping support structures in light-water reactors that exhibited elasto-plastic behavior, and studied the effect of displacement and the vibration frequency on the stiffness and on the energy absorption by testing these models. In the second stage, vibration tests were performed using piping models with support structures on shaking tables. The piping vibration characteristics were clarified by sinusoidal sweep tests and the piping response characteristics by seismic wave vibration tests when the support structures were in an elasto-plastic condition. In the third stage, a general method was developed to evaluate the characteristics of a variety of support structures in the tests. A simplified analysis method was also developed to evaluate the piping seismic response using the piping model test result. To expand the results mentioned above, we also established a new seismic design method of piping systems that allowed support structures to have elasto-plastic behavior. This paper reports the newly developed seismic design method based on the results of experiments conducted under the joint research program of Japanese electric power companies (The Japan Atomic Power Co., Hokkaido EPC, Tohoku EPC, Tokyo EPC, Chubu EPC, Hokuriku EPC, Kansai EPC, Chugoku EPC, Shikoku EPC, Kyushu EPC) and nuclear plant makers (Hitachi Ltd., Toshiba Co., MHI Ltd., HEC Ltd

  11. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1990-12-01

    During the second quarter of 1990 the Finnish nuclear plant units Loviisa 1 and 2 and TVO and II were in commercial operation for most of the time. The feedwater pipe rupture at Loviisa 1 and the resulting inspections and repairs at both Loviisa plant units brought about an outage the overall duration of which was 32 days. The annual maintenance outages of the TVO plant units were arranged during the report period and their combined duration was 31.5 days. Nuclear electricity accounted for 35.3% of the total Finnish electricity production during this quarter. The load factor average of the nuclear power plant units was 83.0%. Three events occurred during the report period which are classified as Level 1 on the International Nuclear Event Scale: feedwater pipe rupture at Loviisa 1, control rod withdrawal at TVO I in a test during an outage when the hydraulic scram system was rendered inoperable and erroneous fuel bundle transfers during control rod drives maintenance at TVO II. Other events during this quarter are classified as Level Zero (Below Scale) on the International Nuclear Event Scale. Occupational radiation doses and external releases of radioactivity were considerably below authorised limits. Only small amounts of nuclides originating in nuclear power plants were detected in samples taken in the vicinity of nuclear power plants

  12. 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

  13. 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

  14. The development of the design method of nuclear piping system supported by elasto-plastic support structures (Part 1)

    International Nuclear Information System (INIS)

    Endo, R.; Murota, M.; Kawahata, J.-I.; Sato, T.; Mekomoto, Y.; Takayama, Y.; Kobayashi, H.; Hirose, J.

    1993-01-01

    The conventional aseismic design method of nuclear piping system is very conservative because of the accumulation of various safety factors in the design process, and nuclear piping systems are thought to have a large safety margin. Considering this situation, we promoted research to further rationalize nuclear power plants by reducing the amount of support structures and reducing the piping seismic response through vibration energy absorption resulting from the elasto-plastic behavior of piping support structures. The research has the following three stages. In the first stage, we select conventional piping support structures in Japanese light-water reactors that exhibit elasto-plastic behavior, and study the displacement dependency and the vibration frequency dependency on the stiffness and the energy absorption by testing their model. In the second stage, we make a piping test model with support structures whose characteristics have already been obtained, and perform vibration tests on a shaking table. In this way, we analyze the piping vibration characteristics by sinusoidal wave sweep tests and the piping response characteristics by seismic wave vibration tests, when the support structures are in an elasto-plastic condition. In the third stage, a general method is developed to evaluate the characteristics of the support structures obtained in the tests and it is applied to the evaluation of the characteristics of general support structures. A simplified analysis method is developed to evaluate the piping seismic response using the piping model test result. To expand the results mentioned above, we are developing a seismic design method of piping systems that allows support structures to have elasto-plastic behaviour. This paper reports the results of experiments conducted under the joint research program of Japanese electric power companies with support elements in the first stage and those with piping models in the second stage

  15. 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…

  16. New design system for nuclear power plant

    International Nuclear Information System (INIS)

    Kakuta, Masataka; Yoshinaga, Toshiaki; Yoshida, Ikuzo; Tokumasu, Shinji.

    1980-01-01

    As for the machine and equipment layout and the piping design for nuclear power plants, the multilateral coordination and study on such factors as functions, installation, radiation exposure and maintenance are required, and the high reliability is demanded. On the other hand, the quantity of things handled is enormous, therefore it is difficult to satisfy completely the above described requirements and to make plant planning which is completely free from the mutual interference of machines, equipments and pipings by the ordinary design with drawings only. Thereupon, the following new device was adopted to the design method for the purposes of improving the quality and shortening the construction period. Namely at the time of designing new plants, the rationalization of plant planning method was attempted by introducing color composite drawings and the technique of model engineering, at the same time, the newly developed design system for pipings was applied with a computer, thus the large accomplishment was able to be obtained regarding the improvement of reliability and others by making the check-up of the propriety. The design procedures of layout and piping, the layout design and general coordination in nuclear power stations with models and color composite drawings and the design system are explained. (Kako, I.)

  17. 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

  18. Structural analysis program of plant piping system. Introduction of AutoPIPE V8i new feature. JSME PPC-class 2 piping code

    International Nuclear Information System (INIS)

    Motohashi, Kazuhiko

    2009-01-01

    After an integration with ADLPipe, AutoPIPE V8i (ver.9.1) became the structural analysis program of plant piping system featured with analysis capability for the ASME NB Class 1 and JSME PPC-Class 2 piping codes including ASME NC Class 2 and ASME ND Class 3. This article described analysis capability for the JSME PPC-Class 2 piping code as well as new general features such as static analysis up to 100 thermal, 10 seismic and 10 wind load cases including different loading scenarios and pipe segment edit function: join, split, reverse and re-order segments. (T. Tanaka)

  19. Development of a software for the ASME code qualification of class-I nuclear piping systems

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Umashankar, C.; Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

    1999-11-01

    In nuclear industry, the designer often comes across the requirements of Class-1 piping systems which need to be qualified for various normal and abnormal loading conditions. In order to have quick design changes and the design reviews at various stages of design, it is quite helpful if a dedicated software is available for the qualification of Class-1 piping systems. BARC has already purchased a piping analysis software CAESAR-II and has used it for the life extension of heavy water plant, Kota. CAESAR-II facilitates the qualification of Class-2 and Class-3 piping systems among others. However, the present version of CAESAR-II does not have the capability to perform stress checks for the ASME Class-1 nuclear piping systems. With this requirement in mind and the prohibitive costs of commercially available software for the Class-1 piping analyses, it was decided to develop a separate software for this class of piping in such a way that the input and output details of the piping from the CAESAR-II software can be made use of. This report principally contains the details regarding development of a software for codal qualification of Class-1 nuclear piping as per ASME code section-III, NB-3600. The entire work was carried out in three phases. The first phase consisted of development of the routines for reading the output files obtained from the CAESAR-II software, and converting them into required format for further processing. In this phase, the nodewise informations available from the CAESAR-II output file were converted into element-wise informations. The second phase was to develop a general subroutine for reading the various input parameters such as diameter, wall thickness, corrosion allowance, bend radius and also to recognize the bend elements based on the bend radius, directly from the input file of CAESAR-II software. The third phase was regarding the incorporation of the required steps for performing the ASME codal checks as per NB-3600 for Class-1 piping

  20. Fabrication of mechanical components and piping design for Brazilian nuclear reactors

    International Nuclear Information System (INIS)

    Deppe, L.O.

    1987-01-01

    The supply of Brazilian equipment and piping design for Angra 2 (and Angra 3 in some cases) have reached an advanced status in spite of the continuous outside difficulties which affect these nuclear power plants. The achieved quality is similar to the quality achieved in foreign countries and the nationalization program foreseen in 1975 is being largely surpassed. In this paper the actual situation is presented as well as the future perspectives. (Author) [pt

  1. 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.)

  2. Piping benchmark problems for the Westinghouse AP600 Standardized Plant

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1997-01-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the Westinghouse AP600 Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the AP600 standard design. It will be required that the combined license licensees demonstrate that their solutions to these problems are in agreement with the benchmark problem set

  3. Trend of field data on pipe wall thinning for BWR power plants

    International Nuclear Information System (INIS)

    Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

    2009-01-01

    Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

  4. 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.)

  5. Seismic analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Halbritter, A.L.

    1984-01-01

    Nuclear Power Plants require exceptional safety guarantees which are reflected in a rigorous control of the employed materials, advanced construction technology, sophisticated methods of analysis and consideration of non conventional load cases such as the earthquake loading. In this paper, the current procedures used in the seismic analysis of Nuclear Power Plants are presented. The seismic analysis of the structures has two objectives: the determination of forces in the structure in order to design it against earthquakes and the generation of floor response spectra to be used in the design of mechanical and electrical components and piping systems. (Author) [pt

  6. Nuclear power plant construction

    International Nuclear Information System (INIS)

    Lima Moreira, Y.M. de.

    1979-01-01

    The legal aspects of nuclear power plant construction in Brazil, derived from governamental political guidelines, are presented. Their evolution, as a consequence of tecnology development is related. (A.L.S.L.) [pt

  7. In-service monitoring of piping and vessels in nuclear power plants. Report on BMU-study SR 2218; Betriebsueberwachung von Rohrleitungen und Behaeltern in Kernkraftwerken. Bericht ueber die BMU-Studie SR 2218

    Energy Technology Data Exchange (ETDEWEB)

    Hofstoetter, P. [Technischer Ueberwachungs-Verein Rheinland e.V., Koeln (Germany); Dittmar, S. [Technischer Ueberwachungs-Verein Nord e.V., Hamburg (Germany); Maile, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Metzner, K.J. [Preussen Elektra AG, Hannover (Germany). Hauptverwaltung

    1998-11-01

    On behalf of the BMU, the authors carried out a study intended to give an overview of and evaluate current practice and methods applied in fatigue monitoring of pipes and vessels of nuclear power plants. The authors on the one hand present fundamentals relating to typical load cases, monitoring and measuring techniques and fatigue computation methods, as well as the relevant legal provisions, and on the other hand describe the monitoring practice, tools and techniques referring to available literature, own experience, and interviews with operators/owners and manufacturers of existing power plants. The locations selected for examinations, measuring techniques, data acquisition and evaluation as well as measuring protocols are explained. The monitoring and inspection methods applied are assessed in a critical review which is the basis of proposals for improvement, relating to the measured data acquisition and especially to data analysis and evaluation. The paper focuses on the essential results of the study, excluding those parts of the study dealing with the calculation of fatigue in components, which will be the subject of another paper to be published. (orig./CB) [Deutsch] Im Auftrag ders BMU erarbeiteten die Verfasser eine Studie zur Darstellung und Bewertung der heute in Kernkraftwerken ueblichen Ermuedungsueberwachung von Rohrleitungen und Behaeltern. Dazu wurden einerseits die Grundlagen zur Beanspruchung, Messtechnik und Ermuedungsberechnung zusammengestellt sowie geltende Vorschriften aufgefuehrt. Andererseits wurde auf der Basis der verfuegbaren Literatur, eigener Erfahrungen sowie einer Befragung von Betreibern und Herstellern der heute ueblichen Anlagen die Messstellen, Messverfahren, Messwerterfassungsanlagen und Auswertungsmethoden sowie bisherige Ergebnisse der Messungen dargestellt. Das derzeitige Vorgehen wurde kritisch bewertet. Daraus resultierten Verbesserungsvorschlaege zur Messwerteerfassung, vor allem aber zur Auswertung und Bewertung der

  8. Integrity of austenitic stainless steel piping welds for nuclear service

    International Nuclear Information System (INIS)

    Canalini, A.; Lopes, L.R.

    1983-01-01

    A criterion applying K 1d concept was developed to determine the fracture mechanics properties of austenitic stainless steel nuclear piping welds. The critical dimensions, lenght and depth, for crack initiation were established and plotted in a chart. This study enables the dimensions of a discontinuity detected in an in-service inspection to be compared to the critical dimensions for crack initiation, and the indication can be judged critical or non-critical for the component. (author) [pt

  9. 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

  10. Nuclear Power Plant Technician

    Science.gov (United States)

    Randall, George A.

    1975-01-01

    The author recognizes a body of basic knowledge in nuclear power plant technoogy that can be taught in school programs, and lists the various courses, aiming to fill the anticipated need for nuclear-trained manpower--persons holding an associate degree in engineering technology. (Author/BP)

  11. 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

  12. Development of an evaluation method for seismic isolation systems of nuclear power facilities. Seismic design analysis methods for crossover piping system

    International Nuclear Information System (INIS)

    Tai, Koichi; Sasajima, Keisuke; Fukushima, Shunsuke; Takamura, Noriyuki; Onishi, Shigenobu

    2014-01-01

    This paper provides seismic design analysis methods suitable for crossover piping system, which connects between seismic isolated building and non-isolated building in the seismic isolated nuclear power plant. Through the numerical study focused on the main steam crossover piping system, seismic response spectrum analysis applying ISM (Independent Support Motion) method with SRSS combination or CCFS (Cross-oscillator, Cross-Floor response Spectrum) method has found to be quite effective for the seismic design of multiply supported crossover piping system. (author)

  13. Model engineering for piping layout of boiling water reactor nuclear station

    International Nuclear Information System (INIS)

    Tsukada, Koji; Uchiyama, Masayuki; Wada, Takanao; Jibu, Noboru.

    1977-01-01

    A nuclear power station is made up of a wide variety of equipment, piping, ventilation ducts, conduits, and cable trays, etc. Even if equipment arrangement and piping layout are carefully planned on drawings, troubles such as interference often occur at field installation. Accordingly, it is thought very useful to make thorough examinations with plastic three-dimensional models in addition to drawings in reducing troubles at field, shortening the construction period, and improving economics. Examination with plastic models offers the following features: (1) It permits visual three-dimensional examination. (2) Group thinking and examination is possible. (3) Troubles due to failure to understand complicated drawings can be reduced drastically. Manufacturing a 1/20 scale model of the reactor building of the Tokai No. 2 Power Station of the Japan Atomic Power Co., Hitachi has performed model engineering-solution of interference troubles related to equipment and piping, securing of work space for in-service inspection (ISI), carry-in/installation of various equipment and piping, and determination of the piping route of which only the starting and terminating points were given under the complicated ambient conditions. Success with this procedure has confirmed that model engineering is an effective technique for future plant engineering. (auth.)

  14. NUCLEAR POWER PLANT

    Science.gov (United States)

    Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

    1963-05-14

    A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

  15. Nuclear power plants maintenance

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    Nuclear power plants maintenance now appears as an important factor contributing to the competitivity of nuclea energy. The articles published in this issue describe the way maintenance has been organized in France and how it led to an actual industrial activity developing and providing products and services. An information note about Georges Besse uranium enrichment plant (Eurodif) recalls that maintenance has become a main data not only for power plants but for all nuclear industry installations. (The second part of this dossier will be published in the next issue: vol. 1 January-February 1989) [fr

  16. Technology of Inspection and Real-time Displacement Monitoring on Critical Pipe for Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyun, Jung Seob; Heo, Jae Sil [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Cho, Sun Young [KLES, Daejeon (Korea, Republic of); Heo, Jeong Yeol; Lee, Seong Kee [Korea South-East Power Co., Seoul (Korea, Republic of)

    2009-10-15

    High temperature steam pipes of thermal power plant are subject to a severe thermal range and usually operates well into the creep range. Cyclic operation of the plant subjects the piping system to mechanical and thermal fatigue damages. Also, poor or malfunctional supports can impose massive loads or stress onto the piping system. In order to prevent the serious damage and failure of the critical piping system, various inspection methods such as visual inspection, computational analysis and on-line piping displacement monitoring were developed. 3-dimensional piping displacement monitoring system was developed with using the aluminum alloy rod and rotary encoder sensors, this system was installed and operated on the high temperature steam piping of 'Y' thermal power plant successfully. It is expected that this study will contribute to the safety of piping system, which could minimize stress and extend the actual life of critical piping.

  17. 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

  18. 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.

  19. Utilization of logistic computer programs in the power plant piping industry

    International Nuclear Information System (INIS)

    Motzel, E.

    1982-01-01

    Starting from the general situation of the power plant piping industry, the utilization of computer programs as well as the specific magnitude of complexity connected with the project realisation, the necessity for using logistic computer programs especially in case of nuclear power plants is explained. The logistic term as well as the logistic data are described. At the example of the nuclear power plant KRB II, Gundremmingen, Block B/C the practical use of such programs is shown. The planning, scheduling and supervision is carried out computer-aided by means of network-technique. The material management, prefabrication, installation including management of certificates for welding and testing activities is planned and controlled by computer programs as well. With the piping systems installed a complete erection work documentation is available which also serves as base for the billing versus the client. The budgeted costs are continuously controlled by means of a cost control program. Summing-up the further development in controlling piping contracts computer-supported is described with regard to software, hardware and the organisation structure. Furthermore the concept of a self-supporting field computer is introduced for the first time. (orig.) [de

  20. 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

  1. Management of manufacture and installation of plant pipings by bar code system

    International Nuclear Information System (INIS)

    Suwa, Minoru

    1995-01-01

    As for the piping system of nuclear power plants, the number of parts is very large, and the mill sheet is attached to each part, therefore, it is necessary to manage them individually, and large man power is required. In order to resolve the delay of mechanization in the factory, bar code system was adopted on full scale. At the time of taking parts out from the store, bar code labels are stuck to all piping parts. By this means, all the processes of manufacture and inspection are managed with a computer, and it is useful for labor saving and the prevention of mistaken input. This system is centering around the system of the progress management for piping manufacture, and is operated by being coupled with respective systems of production design, order and inventory, mill sheet management and installation management. The management of production design, manufacture, inspection and installation is explained. There is the problem of sticking bar code labels again as the labels become dirty or parts pass through coating and pickling processes. The direct carving of bar codes on parts by laser marker was tried, and it was successful for stainless steel, but in carbon steel pipes, it was hard to read. It is desirable to develop the bar codes which endure until the end of plant life. (K.I.)

  2. Nuclear power plant

    International Nuclear Information System (INIS)

    Nishio, Masahide

    1986-01-01

    Purpose: To provide a constitution capable of previously and reliably preventing radioactivity from releasing into the atmosphere upon occurrence of main steam pipe rupture accidents in a main steam tunnel chamber. Constitution: The outer circumference at the penetration portion of a nuclear reactor container is tightly closed and the main steam tunnel chamber has a tightly closed vessel structure, which is cooled by a local cooler during normal operation. The main steam tunnel chamber is in communication with a pressure control chamber by way of a release line and a releaf valve is disposed at the midway of the release line. Upon occurrence of rupture accident to the main steam pipes in the main steam tunnel chamber, while steams are issued from the ruptured portion, they are discharged through the release line to the suppression chamber and condensated. As a result, excess pressure in the main steam tunnel can be prevented and when the rupture accident is detected, the main steam isolation valve is closed rapidly to interrupt the steam feeding, whereby the steam released from the ruptured pipeways is stopped to avoid the radioactivity release to the atmosphere. (Kamimura, M.)

  3. Assessment of LWR piping design loading based on plant operating experience

    International Nuclear Information System (INIS)

    Svensson, P.O.

    1980-08-01

    The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading

  4. Review of nuclear power plant systems

    International Nuclear Information System (INIS)

    Doehler

    1980-01-01

    This presentation starts with a brief description of the Technischer Ueberwachungs-Verein (TUeV) and its main activities in the field of technical assessments. The TUeV-organisation is in general the assessor who performs the review if nuclear power plant systems, structures and equipment. All aspects relating to the safe operation of nuclear power plants are assessed by the TUeV. This paper stresses the review of the design of nuclear power plant systems and structures. It gives an outline on the procedure of an assessment, starting with the regulatory requirements, going into the papers of the applicant and finally ending with the TUeV-appraisal. This procedure is shown using settlement measuring requirements as an example. The review of the design of mechanical structures such as pipes, valves, pump and vessels is shown in detail. (RW)

  5. Construct ability Improvement for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dae Soo; Lee, Jong Rim; Kim, Jong Ku [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The purpose of this study was to identify methods for improving the construct ability of nuclear power plants. This study reviewed several references of current construction practices of domestic and overseas nuclear plants in order to identify potential methods for improving construct ability. The identified methods for improving construct ability were then evaluated based on the applicability to domestic nuclear plant construction. The selected methods are expected to reduce the construction period, improve the quality of construction, cost, safety, and productivity. Selection of which methods should be implemented will require further evaluation of construction modifications, design changes, contract revisions. Among construction methods studied, platform construction methods can be applied through construction sequence modification without significant design changes, and Over the Top construction method of the NSSS, automatic welding of RCL pipes, CLP modularization, etc., are considered to be applied after design modification and adjustment of material lead time. (author). 49 refs., figs., tabs.

  6. Nuclear plant life extension

    International Nuclear Information System (INIS)

    Negin, C.A.

    1989-01-01

    The nuclear power industry's addressing of life extension is a natural trend in the maturation of this technology after 20 years of commercial operation. With increasing emphasis on how plants are operated, and less on how to build them, attention is turning on to maximizing the use of these substantial investments. The first studies of life extension were conducted in the period from 1978 and 1982. These were motivated by the initiation, by the Nuclear Regulatory Commission (NRC), of studies to support decommissioning rulemaking. The basic conclusions of those early studies that life extension is feasible and worth pursuing have not been changed by the much more extensive investigations that have since been conducted. From an engineering perspective, life extension for nuclear plants is fundamentally the same as for fossil plants

  7. Structural mechanics in nuclear power plant

    International Nuclear Information System (INIS)

    Han Liangbi

    1998-01-01

    The main research works in structural mechanics in reactor technology are emphatically introduced. It is completed by structural mechanics engineers at Shanghai Nuclear Research and Design Institute associated with the design and construction problems for Qinshan NPP Unit 1 and Pakistani CHASNUPP. About structural mechanics problem for the containment, the rock and soft soil two different bases are considered. For the later the interaction between soil and structure is carefully studied. About the structural mechanics problem for the equipment and pipings, the three dimensional stress and fracture analyses are studied. For the structural dynamics problem, including flow induced vibration, the response analyses under earthquake and loss coolant accident loadings are studied. For pipings, the leak before break technique has been emphatically introduced. A lot of mathematical models, the used computer codes, analytical calculations and experimental results are also introduced. This is a comprehensive description about structural mechanics problem in pressurized water reactor nuclear power plant

  8. 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).

  9. Nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Yaziz Yunus

    1986-01-01

    A number of issues have to be taken into account before the introduction of any nuclear power plant in any country. These issues include reactor safety (site and operational), waste disposal and, lastly, the decommissioning of the reactor inself. Because of the radioactive nature of the components, nuclear power plants require a different approach to decommission compared to other plants. Until recently, issues on reactor safety and waste disposal were the main topics discussed. As for reactor decommissioning, the debates have been academic until now. Although reactors have operated for 25 years, decommissioning of retired reactors has simply not been fully planned. But the Shippingport Atomic Power Plant in Pennysylvania, the first large scale power reactor to be retired, is now being decommissioned. The work has rekindled the debate in the light of reality. Outside the United States, decommissioning is also being confronted on a new plane. (author)

  10. An underground nuclear power station using self-regulating heat-pipe controlled reactors

    Science.gov (United States)

    Hampel, V.E.

    1988-05-17

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

  11. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    Science.gov (United States)

    Hampel, Viktor E.

    1989-01-01

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

  12. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    International Nuclear Information System (INIS)

    Hampel, V.E.

    1989-01-01

    The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

  13. Diagnosis and on-line displacement monitoring for critical pipe of fossil power plant

    Energy Technology Data Exchange (ETDEWEB)

    Heo, J. S.; Hyun, J. S. [Korea Electric Power Corporation, Seoul (Korea, Republic of); Heo, J. R.; Lee, S. K.; Cho, S. Y. [Korea South-East Power Co., Ltd., Seoul (Korea, Republic of)

    2009-07-01

    High temperature steam pipes of fossil power plant are subject to a severe thermal range and usually operates well into the creep range. Cyclic operation of the plant subjects the piping system to mechanical and thermal fatigue mechanisms and poor or malfunctional support assemblies can impose massive loads or stress onto the piping system. In order to prevent the serious damage and failure of the critical pipe system, various inspection methods such as visual inspection, computational analysis and on-line piping displacement monitoring were developed. 3-Dimensional piping displacement monitoring system was developed with using he aluminum alloy rod and rotary encoder type sensors, this system was installed and operated on the 'Y' fossil power plant successfully. It is expected that this study will contribute to the safety of piping system, which could minimize stress and extend the actual life of critical piping.

  14. Mechanical properties of roll extruded nuclear reactor piping

    International Nuclear Information System (INIS)

    Steichen, J.M.; Knecht, R.L.

    1975-01-01

    The elevated temperature mechanical properties of large diameter (28 inches) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of Type 316H stainless steel piping material used exhibited consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceeded values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050 0 F for times to 10,000 hours. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900 0 F and that for temperatures of 1050 0 F and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations. (U.S.)

  15. Development of robots for nuclear power plants

    International Nuclear Information System (INIS)

    Sasaki, Masayoshi

    1982-01-01

    In nuclear power plants, the reduction of maintenance time, the reduction of radiation exposure and man-power saving are increasingly required. To achieve these purposes, various remote-controlled devices, such as robots in a broad sense, have been earnestly developed. Of these, three machines for replacing, four devices for inspection, two systems for cleaning, and two equipment for processing are tabulated in this paper. Typical eight machines or equipment are briefly introduced, mainly describing their features or characteristics. Those are: a remotely handling machine for control rod drive mechanism, an automatic refueling machine, an automatic ultrasonic flaw detection system replacing for a manually operated testing system for the welded parts of primary cooling system, an automatic cask washing machine for decontamination, a floor-type remote inspection vehicle for various devices operating inside power plants, a monorail-type remote inspection vehicle for inspection in spaces where floor space is short, and a remote-controlled automatic pipe welding machine for welding operations in a radioactive environment such as replacing the piping of primary cooling system. Most of these devices serves for radiation exposure reduction at the same time. Existing nuclear power plant design assumes direct manual maintenance, which limits the introduction of robots. Future nuclear power plants should be designed on the assumption of automatic remote-controlled tools and devices being used in maintenance work. (Wakatsuki, Y.)

  16. Nuclear power plant analyzer

    International Nuclear Information System (INIS)

    Stritar, A.

    1986-01-01

    The development of Nuclear Power Plant Analyzers in USA is described. There are two different types of Analyzers under development in USA, the forst in Idaho and Los Alamos national Lab, the second in brookhaven National lab. That one is described in detail. The computer hardware and the mathematical models of the reactor vessel thermalhydraulics are described. (author)

  17. Nuclear plant scram reduction

    International Nuclear Information System (INIS)

    Wiegle, H.R.

    1986-01-01

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

  18. Nuclear plant license renewal

    International Nuclear Information System (INIS)

    Gazda, P.A.; Bhatt, P.C.

    1991-01-01

    During the next 10 years, nuclear plant license renewal is expected to become a significant issue. Recent Electric Power Research Institute (EPRI) studies have shown license renewal to be technically and economically feasible. Filing an application for license renewal with the Nuclear Regulatory Commission (NRC) entails verifying that the systems, structures, and components essential for safety will continue to perform their safety functions throughout the license renewal period. This paper discusses the current proposed requirements for this verification and the current industry knowledge regarding age-related degradation of structures. Elements of a license renewal program incorporating NRC requirements and industry knowledge including a schedule are presented. Degradation mechanisms for structural components, their significance to nuclear plant structures, and industry-suggested age-related degradation management options are also reviewed

  19. Studies of S-CO{sub 2} Power Plant Pipe Design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minseok; Ahn, Yoonhan; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    Further development of nuclear energy is required to address the global warming issue while overcoming the difficulty of meeting the constantly growing demand of energy. As the nuclear energy does not only reduce the carbon dioxide emission but also attain sufficient and stable electricity supply, this is considered as one of the most clean and sustainable energy sources. The Sodium-cooled Fast Reactor (SFR) is a strong candidate among the next generation nuclear reactors. However, current SFR design may face difficulty in public acceptance due to the potential hazard from sodium-water reaction (SWR) when the current conventional steam Rankine cycle is utilized as a power conversion system for SFR. In order to eliminate SWR, the Supercritical CO{sub 2} (S-CO{sub 2}) cycle has been proposed. Although many S-CO{sub 2} cycle concepts are being suggested by many research organizations, pipe selection criteria for S-CO{sub 2} cycle are one of the areas that are not clearly established. As one of the most important parts of the plant design is economical fluid transfer, this paper will discuss how to select a suitable pipe for the S-CO{sub 2} power plant compared to steam Rankine cycle. The main advantages of S-CO{sub 2} cycle are: prevention of no SWR by changing the working fluid, relatively high efficiency with 450∼750 .deg. C turbine inlet temperature, physically compact size. Additional study for larger system such as 300MW class system in MIT report will be conducted. From the preliminary estimation when the S-CO{sub 2} system becomes large than the pipe diameter may exceed the current ASME standard. This means that more innovative approach will be needed for the S-CO{sub 2} pipe design. To economically design the pipe of S-CO{sub 2} recompressing cycle, optimal flow velocity for S-CO{sub 2} that can be obtained through the process engineering should exist. Although the Ronald W. Capps equation offers an optimal flow velocity while considering safety, capital

  20. Nuclear plant safety

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The four-member New York Power Pool Panel concluded that, for a number of reasons, no nuclear power plant in New York State is prone to the type of accident that occurred at Three Mile Island (TMI). The Panel further concluded that changes in operating practices, both regulatory and voluntary, and heightened sensitivity to reactor-core-cooling requirements will substantially reduce the chances for another such accident anywhere. Panel members found that New York State utilities have taken a responsible attitude with regard to requirements set forth by the Nuclear Regulatory Commission (NRC) as a result of the TMI accident. In a cover letter that accompanied the report to Federal and New York state officials, New York Power Pool Executive Committee Chairman Francis E. Drake, Jr. expressed hope that the report will alleviate public fears of nuclear reactors and promote wider acceptance of nuclear energy as an economic and safe means of power production. 17 references

  1. Nuclear plants - military hostages

    International Nuclear Information System (INIS)

    Ramberg, B.

    1986-01-01

    Recent events suggest that nuclear reactors could make tempting military or terrorist targets. Despite the care with which most reactors are built, studies document their vulnerability to willful destruction through disruption of coolant mechanisms both inside and outside the containment building. In addition to reactors, such nuclear support facilities as fuel fabrication, reprocessing, and waste storage installations may be attractive military targets. A nuclear bomb which exploded in the vicinity of a reactor could increase its lethal effects by one-third. The implications of this is vulnerability for Middle East stability as well as to other volatile regions. The author suggests several avenues for controlling the dangers: international law, military and civil defense, facility siting, increasing plant safety, and the international management of nuclear energy. 21 references

  2. Management of nuclear power plants lifetime

    International Nuclear Information System (INIS)

    Hutin, J.P.

    2006-01-01

    The factors influencing the management of the service life of nuclear power plants can be of various types and the 'heaviest' ones have to be managed through robust and explicit approaches involving all actors. However, the mastery of the service life starts with the mastery of the technical problems, in particular the physical aging of the facilities. This mastery requires to foresee and anticipate the problems and thus a good understanding of the phenomena involved. This article presents: 1 - the general problem of service life management: lifetime concept, situation of French power plants, service life management policy; 2 - aging mechanisms: embrittlement of steel under irradiation, swelling of materials, thermal aging, fatigue, stress corrosion, aqueous corrosion of metals, corrosion-erosion, mechanisms of concrete degradation, mechanisms of elastomers and polymers degradation, wear; 3 - non-replaceable parts: reactor vessel, containment building; 4 - replaceable parts: cables, instrumentation and control system, core internals, primary loop piping, auxiliary primary piping, pressurizer, primary pump, steam generator tubes, other Ni-Cr-Fe alloy parts, secondary loop piping, turbine, alternator; 5 - non-technical aspects: perenniality of the industrial support, evolution of safety requirements, public acceptance, economical aspects, knowledge and information systems; 6 - situation in foreign countries: status of the world nuclear park, lifetime notion in foreign countries, situation in the USA. (J.S.)

  3. Pulsed nuclear power plant

    International Nuclear Information System (INIS)

    David, C.V.

    1986-01-01

    This patent describes a nuclear power plant. This power plant consists of: 1.) a cavity; 2.) a detonatable nuclear device in a central region of the cavity; 3.) a working fluid inside of the cavity; 4.) a method to denote a nuclear device inside of the cavity; 5.) a mechanical projection from an interior wall of the cavity for recoiling to absorb a shock wave produced by the detonation of the nuclear device and thereby protecting the cavity from damage. A plurality of segments defines a shell within the cavity and a plurality of shock absorbers, each connecting a corresponding segment to a corresponding location on the wall of the cavity. Each of these shock absorbers regulate the recoil action of the segments; and 6.) means for permitting controlled extraction of a quantity of hot gases from the cavity produced by the vaporization of the working fluid upon detonation of the nuclear device. A method of generating power is also described. This method consists of: 1.) introducing a quantity of water in an underground cavity; 2.) heating the water in the cavity to form saturated steam; 3.) detonating a nuclear device at a central location inside the cavity; 4.) recoiling plate-like elements inside the cavity away from the central location in a mechanically regulated and controlled manner to absorb a shock wave produced by the nuclear device detonation and thereby protect the underground cavity against damage; 5.) extracting a quantity of superheated steam produced by the detonation of the nuclear device; and 6.) Converting the energy in the extracted superheated steam into electrical power

  4. Design of a nuclear steam reforming plant

    International Nuclear Information System (INIS)

    Malherbe, J.

    1980-01-01

    The design of a plant for the steam reforming of methane using a High Temperature Reactor has been studied by CEA in connection with the G.E.G.N. This group of companies (CEA, GAZ DE FRANCE, CHARBONNAGES DE FRANCE, CREUSOT-LOIRE, NOVATOME) is in charge of studying the feasibility of the coal gasification process by using a nuclear reactor. The process is based on the hydrogenation of the coal in liquid phase with hydrogen produced by a methane steam reformer. The reformer plant is fed by a pipe of natural gas or SNG. The produced hydrogen feeds the gasification plant which could not be located on the same site. An intermediate hydrogen storage between the two plants could make the coupling more flexible. The gasification plant does not need a great deal of heat and this heat can be satisfied mostly by internal heat exchanges

  5. Beloyarsk Nuclear Power Plant

    International Nuclear Information System (INIS)

    1997-01-01

    The Beloyarsk Nuclear Power Plant (BNPP) is located in Zarechny, approximately 60 km east of Ekaterinberg along the Trans-Siberian Highway. Zarechny, a small city of approximately 30,000 residents, was built to support BNPP operations. It is a closed city to unescorted visitors. Residents must show identification for entry. BNPP is one of the first and oldest commercial nuclear power plants in Russia and began operations in 1964. As for most nuclear power plants in the Russian Federation, BNPP is operated by Rosenergoatom, which is subordinated to the Ministry of Atomic Energy of the Russian Federation (Minatom). BNPP is the site of three nuclear reactors, Units 1, 2, and 3. Units 1 and 2, which have been shut-down and defueled, were graphite moderated reactors. The units were shut-down in 1981 and 1989. Unit 3, a BN-600 reactor, is a 600 MW(electric) sodium-cooled fast breeder reactor. Unit 3 went on-line in April 1980 and produces electric power which is fed into a distribution grid and thermal power which provides heat to Zarechny. The paper also discusses the SF NIKIET, the Sverdiovsk Branch of NIKIET, Moscow, which is the research and development branch of the parent NIKEIT and is primarily a design institute responsible for reactor design. Central to its operations is a 15 megawatt IVV research reactor. The paper discusses general security and fissile material control and accountability at these two facilities

  6. Margins for an in-plant piping system under dynamic loading

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. 4 refs., 6 tabs

  7. Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Booker, C.P.; Turner, M.R.; Spore, J.W.

    1986-01-01

    The Nuclear Plant Data Bank (NPDB) is being developed at the Los Alamos National Laboratory to assist analysts in the rapid and accurate creation of input decks for reactor transient analysis. The NPDB will reduce the time and cost of the creation or modification of a typical input deck. This data bank will be an invaluable tool in the timely investigation of recent and ongoing nuclear reactor safety analysis. This paper discusses the status and plans for the NPDB development and describes its anticipated structure and capabilities

  8. Report of the U.S. Nuclear Regulatory Commission Piping Review Committee. Summary and evaluation of historical strong-motion earthquake seismic response and damage to aboveground industrial piping

    International Nuclear Information System (INIS)

    1985-04-01

    The primary purpose of this report is to collect in one reference document the observation and experience that has been developed with regard to the seismic behavior of aboveground, building-supported, industrial-type process piping (similar to piping used in nuclear power plants) in strong-motion earthquakes. The report will also contain observations regarding the response of piping in strong-motion experimental tests and appropriate conclusions regarding the behavior of such piping in large earthquakes. Recommendations are included covering the future design of such piping to resist earthquake motion damage based on observed behavior in large earthquakes and simulated shake table testing. Since available detailed data on the behavior of aboveground (building-supported) piping are quite limited, this report will draw heavily on the observations and experiences of experts in the field. In Section 2 of this report, observed earthquake damage to aboveground piping in a number of large-motion earthquakes is summarized. In Section 3, the available experience from strong-motion testing of piping in experimental facilities is summarized. In Section 4 are presented some observations that attempt to explain the observed response of piping to strong-motion excitation from actual earthquakes and shake table testing. Section 5 contains the conclusions based on this study and recommendations regarding the future seismic design of piping based on the observed strong-motion behavior and material developed for the NPC Piping Review Committee. Finally, in Section 6 the references used in this study are presented. It should be understood that the use of the term piping in this report, in general, is limited to piping supported by building structures. It does not include behavior of piping buried in soil media. It is believed that the seismic behavior of buried piping is governed primarily by the deformation of the surrounding soil media and is not dependent on the inertial response

  9. Survey of a wireless NDT service for a nuclear piping wall thinning defect

    International Nuclear Information System (INIS)

    Choi, Yoo Rark; Lee, Jae Cheol

    2008-01-01

    The wireless sensor network has been issued for several years. The nuclear power plants all around world have adapted many kinds of sensor technologies for inspections and diagnostics of main instruments. Even though wireless sensor is more useful than wired sensor, wireless sensor based applications haven't been used in nuclear power plants because of the authorization of a jamming, an electromagnetic interference and so on. A wireless sensor uses a battery for its operations, but this battery can't be used for a long haul. It causes a battery change problem. There aren't any wireless sensor based NDT for a piping wall thinning part. We will describe a method of how to develop it in this paper

  10. Ardennes nuclear power plant

    International Nuclear Information System (INIS)

    1974-12-01

    The SENA nuclear power plant continued to operate, as before, at authorized rated power, namely 905MWth during the first half year and 950MWth during the second half year. Net energy production:2028GWh; hours phased to the line: 7534H; availability factor: 84%; utilization factor: 84%; total shutdowns:19; number of scrams:10; cost per KWh: 4,35 French centimes. Overall, the plant is performing very satisfactory. Over the last three years net production has been 5900GWh, corresponding to in average utilization factor of 83%

  11. Nuclear power plant

    International Nuclear Information System (INIS)

    Orlov, V.V.; Rineisky, A.A.

    1975-01-01

    The invention is aimed at designing a nuclear power plant with a heat transfer system which permits an accelerated fuel regeneration maintaining relatively high initial steam values and efficiency of the steam power circuit. In case of a plant with three circuits the secondary cooling circuit includes a steam generator with preheater, evaporator, steam superheater and intermediate steam superheater. At the heat supply side the latter is connected with its inlet to the outlet of the evaporator and with its outlet to the low-temperature side of the secondary circuit

  12. 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.)

  13. Third generation nuclear plants

    Science.gov (United States)

    Barré, Bertrand

    2012-05-01

    After the Chernobyl accident, a new generation of Light Water Reactors has been designed and is being built. Third generation nuclear plants are equipped with dedicated systems to insure that if the worst accident were to occur, i.e. total core meltdown, no matter how low the probability of such occurrence, radioactive releases in the environment would be minimal. This article describes the EPR, representative of this "Generation III" and a few of its competitors on the world market.

  14. Water hammer and cavitational hammer in process plant pipe systems

    International Nuclear Information System (INIS)

    Dudlik, A.; Schoenfeld, S.B.H.; Hagemann, O.; Fahlenkamp, H.

    2003-01-01

    Fast acting valves are often applied for quick safety shut-down of pipelines for liquids and gases in the chemical and petrochemical industry as well as in power plants and state water supplies. The fast deceleration of the liquid leads to water hammer upstream the valve and to cavitational hammer downstream the fast closing valve. The valve characteristics given by manufacturers are usually measured at steady state flow conditions of the liquid. In comparison, the dynamic characteristics depend on the initial liquid velocity, valve closing velocity, the absolute pipe pressure and the pipe geometry. Fraunhofer UMSICHT conducts various test series examining valve dynamic characteristics in order of the dynamic analysis of pressure surges in fast closing processes. Therefore a test rig is used which consists of two pipelines of DN 50 and DN 100 with an approximate length of 230 m each. In this paper the results of performed pressure surge experiments with fast closing and opening valves will be compared to calculations of commercial software programs such as MONA, FLOWMASTER 2. Thus the calculation software for water supply, power plants oil and gas and chemical industry can be permanently improved. (orig.)

  15. Update of foundation design modifications of data cables and piping in nuclear power plants in operation; Actualizacion de modificaciones de sieno de bases de datos de cables y conducciones en centrales nucleares en operacion

    Energy Technology Data Exchange (ETDEWEB)

    Perez Pereira, J.

    2013-07-01

    The scope of this application is the manage the life cycle of cables electrical and pipes of cables in Trillo NPP. The application is integrated in a configuration Control system, so both cables and conduits become elements of configuration and management of life and history associated with the of the relevant modifying documents. The guarantees criteria of physical separation of wires for jobs and for independent networks designed according to the redundancy of the Central System.

  16. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

    International Nuclear Information System (INIS)

    Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

    1981-08-01

    This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations

  17. 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)

  18. Nuclear Power Plant 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Again this year, our magazine presents the details of the conference on Spanish nuclear power plant operation held in February and that was devoted to 1996 operating results. The Protocol for Establishment of a New Electrical Sector Regulation that was signed last December will undoubtedly represent a new challenge for the nuclear industry. By clearing stating that current standards of quality and safety should be maintained or even increased if possible, the Protocol will force the Sector to improve its productivity, which is already high as demonstrated by the results of the last few years described during this conference and by recent sectorial economic studies. Generation of a nuclear kWh that can compete with other types of power plants is the new challenge for the Sector's professionals, who do not fear the new liberalization policies and approaching competition. Lower inflation and the resulting lower interest rates, apart from being representative indices of our economy's marked improvement, will be very helpful in facing this challenge. (Author)

  19. 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)

  20. 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.

  1. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Bain, R.A.; Testa, M.F.

    1994-01-01

    Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

  2. Siting nuclear power plants

    International Nuclear Information System (INIS)

    Yellin, J.; Joskow, P.L.

    1980-01-01

    The first edition of this journal is devoted to the policies and problems of siting nuclear power plants and the question of how far commercial reactors should be placed from urban areas. The article is divided into four major siting issues: policies, risk evaluation, accident consequences, and economic and physical constraints. One concern is how to treat currently operating reactors and those under construction that were established under less-stringent criteria if siting is to be used as a way to limit the consequences of accidents. Mehanical cost-benefit analyses are not as appropriate as the systematic use of empirical observations in assessing the values involved. Stricter siting rules are justified because (1) opposition because of safety is growing: (2) remote siting will make the industry more stable; (3) the conflict is eliminated between regulatory policies and the probability basis for nuclear insurance; and (4) joint ownership of utilities and power-pooling are increasing. 227 references, 7 tables

  3. Nuclear power plant disasters

    International Nuclear Information System (INIS)

    Trott, K.R.

    1979-01-01

    The possibility of a nuclear power plant disaster is small but not excluded: in its event, assistance to the affected population mainly depends on local practitioners. Already existing diseases have to be diagnosed and treated; moreover, these physicians are responsible for the early detection of those individuals exposed to radiation doses high enough to induce acute illness. Here we present the pathogenesis, clinical development and possible diagnostic and therapeutical problems related to acute radiation-induced diseases. The differentiation of persons according to therapy need and prognosis is done on the sole base of the clinical evidence and the peripheral blood count. (orig.) [de

  4. Garigliano nuclear power plant

    International Nuclear Information System (INIS)

    1976-03-01

    During the period under review, the Garigliano power station produced 1,028,77 million kWh with a utilization factor of 73,41% and an availability factor of 85,64%. The disparity between the utilization and availability factors was mainly due to a shutdown of about one and half months owing to lack of staff at the plant. The reasons for nonavailability (14.36%) break down as follows: nuclear reasons 11,49%; conventional reasons 2,81%; other reasons 0,06%. During the period under review, no fuel replacements took place. The plant functioned throughout with a single reactor reticulation pump and resulting maximum available capacity of 150 MWe gross. After the month of August, the plant was operated at levels slightly below the maximum available capacity in order to lengthen the fuel cycle. The total number of outages during the period under review was 11. Since the plant was brought into commercial operation, it has produced 9.226 million kWh

  5. Survey of strong motion earthquake effects on thermal power plants in California with emphasis on piping systems. Volume 2, Appendices

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-11-01

    Volume 2 of the ''Survey of Strong Motion Earthquake Effects on Thermal Power Plants in California with Emphasis on Piping Systems'' contains Appendices which detail the detail design and seismic response of several power plants subjected to strong motion earthquakes. The particular plants considered include the Ormond Beach, Long Beach and Seal Beach, Burbank, El Centro, Glendale, Humboldt Bay, Kem Valley, Pasadena and Valley power plants. Included is a typical power plant piping specification and photographs of typical power plant piping specification and photographs of typical piping and support installations for the plants surveyed. Detailed piping support spacing data are also included

  6. Surface crack behavior in socket weld of nuclear piping under fatigue loading condition

    International Nuclear Information System (INIS)

    Choi, Y.H.; Kim, J.S.; Choi, S.Y.

    2005-01-01

    The ASME B and PV Code Sec. III allows the socket weld for the nuclear piping in spite of the weakness on the weld integrity. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because many failures and leaks have been reported in the socket weld. OPDE (OECD Piping Failure Data Exchange) database lists 108 socket weld failures among 2,399 nuclear piping failure cases during 1970 to 2001. Eleven failures in the socket weld were also reported in Korean NPPs. Many failure cases showed that the root cause of the failure is the fatigue and the gap requirement for the socket weld given in ASME Code was not satisfied. The purpose of this paper is to evaluate the fatigue crack behavior of a surface crack in the socket weld under fatigue loading condition considering the gap effect. Three-dimensional finite element analysis was performed to estimate the fatigue crack behavior of the surface crack. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P=0 to 15.51 MPa, and the thermal transient ranging from T=25 C to 288 C were considered. The results are as follows; 1) The socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME operation and maintenance (OM) Code. 2) The effect of pressure or temperature transient load on the socket weld integrity is not significant. 3) No-gap condition gives very high possibility of the crack initiation at the socket weld under vibration loading condition. 4) For the specific systems having the vibration condition to exceed the requirement in the ASME Code OM and/or the transient loading condition from P=0 and T=25 C to P=15.51 MPa and T=288 C, radiographic examination to examine the gap during the construction stage is recommended. (orig.)

  7. Nuclear power plant emergency preparedness

    International Nuclear Information System (INIS)

    2005-01-01

    The guide sets forth detailed requirements on how the licensee of a nuclear power plant shall plan, implement and maintain emergency response arrangements. The guide is also applied to nuclear material and nuclear waste transport in situations referred to in guide YVL 6.5. Requirements on physical protection are presented in a separate guide of Finnish Radiation and Nuclear Safety Authority (STUK)

  8. Prediction on corrosion rate of pipe in nuclear power system based on optimized grey theory

    International Nuclear Information System (INIS)

    Chen Yonghong; Zhang Dafa; Chen Dengke; Jiang Wei

    2007-01-01

    For the prediction of corrosion rate of pipe in nuclear power system, the pre- diction error from the grey theory is greater, so a new method, optimized grey theory was presented in the paper. A comparison among predicted results from present and other methods was carried out, and it is seem that optimized grey theory is correct and effective for the prediction of corrosion rate of pipe in nuclear power system, and it provides a fundamental basis for the maintenance of pipe in nuclear power system. (authors)

  9. 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

  10. Application of complex engineering solutions through advanced composite innovation (for repair of degraded buried pipe at Vandellos II Nuclear Power Plant); Reparacion de tuberias de un sistema de servicios no esenciales con recubrimiento interno de fibra de carbono

    Energy Technology Data Exchange (ETDEWEB)

    Bueno, J. M.; Raji, B. B.

    2011-07-01

    This technical presentation is focused on introducing an engineering solution approach and identification of sensitivity of applications of advanced carbon fiber in a pressurized wet environment: Engineering design, quality assurance of installation, inspection, and a comprehensive testing program to validate and bench mark the design data and compliance with code requirements in nuclear power plants.

  11. 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

  12. 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.)

  13. Generation of cross section data of heat pipe working fluids for compact nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Slewinski, Anderson; Ribeiro, Guilherme B. [Instituto Tecnológico de Aeronáutica (ITA), São José dos Campos, SP (Brazil); Caldeira, Alexandre D., E-mail: anderson_sle@live.com, E-mail: alexdc@ieav.cta.br, E-mail: gbribeiro@ieav.cta.br [Instituto de Estudos Avançados (IEAv), São José dos Campos, SP (Brazil). Divisão de Energia Nuclear

    2017-07-01

    For compact nuclear power plants, such as the nuclear space propulsion proposed by the TERRA project, aspects like mass, size and efficiency are essential drivers that must be managed during the project development. Moreover, for high temperature reactors, the use of liquid metal heat pipes as the heat removal mechanism provides some important advantages as simplicity and reliability. Considering these aforementioned aspects, this paper aims the development of the procedure necessary to calculate the microscopic absorption cross section data of several liquid metal to be used as working fluids with heat pipes; which will be later compared with the given data from JEF Report ⧣14. The information necessary to calculate the cross section data will be obtained from the latest ENDF library version. The NJOY system will be employed with the following modules: RECONR, BROADR, UNRESR and GROUPR, using the same specifications used to calculate the cross section data encountered in the JEF Report ⧣14. This methodology allows a comparison with published values, verifying the procedure developed to calculate the microscopic absorption cross section for selected isotopes using the TERRA reactor spectrum. Liquid metals isotopes of Sodium (Na), Lithium (Li), Thallium (TI) and Mercury (Hg) are part of this study. (author)

  14. Task force activity to take the effect of elastic-plastic behaviour into account on the seismic safety evaluation of nuclear piping systems

    International Nuclear Information System (INIS)

    Nakamura, Izumi; Shiratori, Masaki; Morishita, Masaki; Otani, Akihito; Shibutani, Tadahito

    2015-01-01

    According to investigations of several nuclear power plants (NPPs) hit by actual seismic events and a number of experimental researches on the failure behavior of piping systems under seismic loads, it is recognized that piping systems used in NPPs include a large seismic safety margin until boundary failure. Since the stress assessment based on the elastic analysis does not reflect actual seismic capability of piping systems including plastic region, it is necessary to develop a rational procedures to estimate the elastic-plastic behavior of piping systems under a large seismic load. With the aim of establishing a procedure that takes into account the elastic-plastic behavior effect in the seismic safety estimation of nuclear piping systems, a task force activity has been planned. Through the activity, the authors intend to establish guidelines to estimate the elastic-plastic behavior of piping systems rationally and conservatively, and to provide new rational seismic safety criteria taking the effect of elastic-plastic behavior into account. As the first step of making out the analysis guideline, benchmark analyses are conducted for a pipe element test and a piping system test. In this paper, the outline of the research activity and the preliminary results of benchmark analyses are described. (author)

  15. 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.)

  16. Wuergassen nuclear power plant

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    The decision of the Federal Court of Administration concerns an application for immediate decommissioning of a nuclear power plant (Wuergassen reactor): The repeal of the permit granted. The decision dismisses the appeal for non-admission lodged by the plaintiffs against the ruling of the Higher Court of Administration (OVG) of North-Rhine Westphalia of December 19th 1988 (File no. 21 AK 8/88). As to the matter in dispute, the Federal Court of Administration confirms the opinion of the Higher Court of Administration. As to the headnotes, reference can be made to that decision. Federal Court of Administration, decision of April 5th 1989 - 7 B 47.89. Lower instance: OVG NW, Az.: 21 AK 8/88. (orig./RST) [de

  17. Nuclear power plant

    International Nuclear Information System (INIS)

    Uruma, Hiroshi

    1998-01-01

    In the first embodiment of the present invention, elements less activated by neutrons are used as reactor core structural materials placed under high neutron irradiation. In the second embodiment of the present invention, materials less activated by neutrons when corrosive materials intrude to a reactor core are used as structural materials constituting portions where corrosion products are generated. In the third embodiment, chemical species comprising elements less activated by neutrons are used as chemical species to be added to reactor water with an aim of controlling water quality. A nuclear power plant causing less radioactivity can be provided by using structural materials comprising a group of specific elements hardly forming radioactivity by activation of neutrons or by controlling isotope ratios. (N.H.)

  18. 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

  19. Pipe support for use in a nuclear system

    International Nuclear Information System (INIS)

    Pollono, L.P.; Mello, R.M.

    1976-01-01

    Description is given of a vertical pipe support system. It comprises a tubular pipe support structure having the same inside diameter and the same wall thickness as the pipe, the pipe support structure having a generally triangularly shaped extension formed integral with and extending circumferentially around its outward side, the bottom side of this extension generally forming a ledge; an annular load-bearing insulation formed adjacent to the extension; means for clamping the load-bearing insulation to extension; and means for providing constant vertical support to means for clamping [fr

  20. Reliability estimation of structures under stochastic loading—A case study on nuclear piping

    International Nuclear Information System (INIS)

    Hari Prasad, M.; Rami Reddy, G.; Dubey, P.N.; Srividya, A.; Verma, A.K.

    2013-01-01

    Highlights: ► Structures are generally subjected to different types of loadings. ► One such type of loading is random sequence and has been treated as a stochastic fatigue loading. ► In this methodology both stress amplitude and number of cycles to failure have been considered as random variables. ► The methodology has been demonstrated with a case study on nuclear piping. ► The failure probability of piping has been estimated as a function of time. - Abstract: Generally structures are subjected to different types of loadings throughout their life time. These loads can be either discrete in nature or continuous in nature and also these can be either stationary or non stationary processes. This means that the structural reliability analysis not only considers random variables but also considers random variables which are functions of time, referred to as stochastic processes. A stochastic process can be viewed as a family of random variables. When a structure is subjected to a random loading, based on the stresses developed in the structure and failure criteria the failure probability can be estimated. In practice the structures are designed with higher factor of safety to take care of such random loads. In such cases the structure will fail only when the random loads are cyclic in nature. In traditional reliability analysis, the variation in the load is treated as a random variable and to account for the number of occurrences of the loading the concept of extreme value theory is used. But with this method one is neglecting the damage accumulation that will take place from one loading to another loading. Hence, in this paper, a new way of dealing with these types of problems has been discussed by using the concept of stochastic fatigue loading. The random loading has been considered as earthquake loading. The methodology has been demonstrated with a case study on nuclear power plant piping.

  1. 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.

  2. 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

  3. 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

  4. The modularization construction of piping system installation in AP1000 plant

    International Nuclear Information System (INIS)

    Lu Song; Wang Yuan; Wei Junming

    2012-01-01

    Modularization construction is the main technique used in AP1000 plants, the piping Modularization installation will impact directly to the module construction as the important part of the Modularization construction. After the piping system has took the modularization design in AP1000 plants, some installation works of piping system has moved from the site to fabrication shop. With improving the construction quality and minimizing the time frame of project, the critical paths can be optimized. This paper has analyzed the risk and challenge that met during the modularization construction period of piping systems though introducing the characteristic of modularization construction for AP1000 piping systems, and get construction experiences from the First AP1000 plants in the world, then it will be the firmly basics for the wide application of modularization construction in the future. (authors)

  5. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1990-03-01

    In the Quarterly Reports on the operation of the Finnish nuclear power plants such incidents and observations are described relating to nuclear and radiation safety which the regulatory body, the Finnish Centre for Radiation and Nuclear Safety, considers safety-related. During the third quarter of 1989 the Finnish nuclear power plant units Loviisa 1 and 2 and TVO I and II were in commercial operation for most of the time. Nuclear electricity accounted for 39.0% of the total Finnish electricity production in this quarter. The load factor average of the nuclear power plant units was 78.9%. At Loviisa 1, two holes were found in the feedwater distributor of one steam generator. Corresponding wall thinning corrosion was also detected in the walls of two other distributors. The holes were found on the feedwater distributor upper surface in the joint of the secondary circuit feedwater pipe. One hole was about 20 mm x 50 mm in size and the other was a pit hole ca 5 mm in diameter. Metal power had entered the primary circuit at TVO I. This was observed during a post-scram plant start-up. Several control rod drive units had become jammed so tight that control rod withdrawal failed. Metal powder did not hamper reactor scram under the prevailing circumstances because the drive units are prone to jamming only after a control rod is almost fully inserted and because the forces which insert a control rod by various means (electrical, hydraulic) are 6-8 fold compared with the withdrawing force

  6. Application of PHADEC method for the decontamination of radioactive steam piping components of Caorso plant

    International Nuclear Information System (INIS)

    Lo Frano, R.; Aquaro, D.; Fontani, E.; Pilo, F.

    2014-01-01

    Highlights: • Application of PHADEC chemical off-line methodology. • Decontamination of radioactive steam piping components of Caorso turbine building. • Experimental characterization of metallic components, e.g., by SEM analysis. • Measure of the efficiency of treatment by means of the reduction of activity and vs. the treatment time. • Minimization of secondary waste produced during decontamination activity of Caorso BWR plant. - Abstract: The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e., taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the

  7. 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.

  8. Acoustic valve leak detection in nuclear plants

    International Nuclear Information System (INIS)

    Dimmick, J.G.; Dickey, J.W.

    1983-01-01

    Internal valve leakage is a hidden energy loss and can cause or prolong a forced outage. Recent advances in acoustic detection of internal valve leakage have reduced piping system maintenance costs, unnecessary downtime, and energy waste. Extremely short payback periods have been reported by plants applying this technology to preventive maintenance, troubleshooting, energy conservation and outage planning. Sensors temporarily attached to the outside of valves and connected to the instruments detect ultrasonic acoustic emissions which are characteristic of internal valve leakage. Since the sensors are attached to the outside of the valves, the time and expense of dismantling the valves or removing them from the systems are eliminated. This paper describes the instrumentation and specific applications to nuclear plant valves, including independent verification of initial findings. Guidelines for potential users, including instrumentation selection, training requirements, application planning, and the choice of in-house versus contract services are discussed

  9. 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.)

  10. 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.

  11. Development of non-destructive diagnosis technology for pipe internal in thermal power plants based on robotics

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seungho; Kim, Changhoi; Seo, Yongchil; Lee, Sunguk; Jung, Seungho; Jung, Seyoung

    2011-11-15

    The Pipelines of power plants may have tiny crack by corrosion. Pipe safety inspection should be performed periodically and non-periodically to ensure their safety and integrity. It is difficult to inspection pipes inside defect since pipes of power plant is covered thermal insulation material. Normally pipes inspection was performed part of pipes on outside. A mobile robot was developed for the inspection of pipe of 100 mm inside diameter. The robot is adopted screw type drive mechanism in order to move vertical, horizontal pipes inside. The multi-laser and camera module, which is mounted in front of the robot, captures a sequence of 360 degree shapes of the inner surface of a pipe. The 3D inner shape of pipe is reconstructed from a multi laser triangulation techniques for the inspection of pipes.

  12. 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.

  13. Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

    International Nuclear Information System (INIS)

    Jong, Rudolf Peter de

    2004-01-01

    Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

  14. 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

  15. 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)

  16. Balance of Plant Requirements for a Nuclear Hydrogen Plant

    Energy Technology Data Exchange (ETDEWEB)

    Bradley Ward

    2006-04-01

    This document describes the requirements for the components and systems that support the hydrogen production portion of a 600 megawatt thermal (MWt) Next Generation Nuclear Plant (NGNP). These systems, defined as the "balance-of-plant" (BOP), are essential to operate an effective hydrogen production plant. Examples of BOP items are: heat recovery and heat rejection equipment, process material transport systems (pumps, valves, piping, etc.), control systems, safety systems, waste collection and disposal systems, maintenance and repair equipment, heating, ventilation, and air conditioning (HVAC), electrical supply and distribution, and others. The requirements in this document are applicable to the two hydrogen production processes currently under consideration in the DOE Nuclear Hydrogen Initiative. These processes are the sulfur iodide (S-I) process and the high temperature electrolysis (HTE) process. At present, the other two hydrogen production process - the hybrid sulfur-iodide electrolytic process (SE) and the calcium-bromide process (Ca-Br) -are under flow sheet development and not included in this report. While some features of the balance-of-plant requirements are common to all hydrogen production processes, some details will apply only to the specific needs of individual processes.

  17. 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.)

  18. Nuclear power plants

    International Nuclear Information System (INIS)

    Usui, Eizo.

    1981-01-01

    Purpose: To prevent boiling of saturated water in the drain tank of a humidity separator by charging cooling water in the drain tank upon power decrease of a turbine. Constitution: Saturated water is separated from high pressure turbine exhausts in a humidity separator and stored in a drain tank. The saturated water in the drain tank is controlled to a constant level and the excess water is sent to a condensator and a feedwater heater. A cooling water feed pipe is branched as a cooling water feed pipe from the exhaust side of a reactor feedwater pump and connected by way of a closing valve to a spray nozzle provided in the drain tank. While the closing valve is usually closed to keep the water level constant in the drain tank, the closing valve is opened upon sudden decrease in the turbine power to charge the condensates by way of the cooling water feed pipe to the drain tank. Thus, the saturated water is mixed with the dondensates and the temperature is lowered to prevent boiling of the saturated water. (Kawakami, Y.)

  19. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    Wood, R.S.

    1991-07-01

    This report indicates percentage ownership of commercial nuclear power plants by utility companies. The report includes all plants operating, under construction, docketed for NRC safety and environmental reviews, or under NRC antitrust review, but does not include those plants announced but not yet under review or those plants formally cancelled. Part 1 of the report lists plants alphabetically with their associated applicants or licensees and percentage ownership. Part 2 lists applicants or licensees alphabetically with their associated plants and percentage ownership. Part 1 also indicates which plants have received operating licenses (OLS)

  20. Nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Hidehiro; Oya, Takashi

    1996-11-05

    The present invention provides a highly safe light water-cooled type nuclear power plant capable of reducing radiation dose by suppressing deposition of activated corrosion products by a simple constitution. Namely, equipments and pipelines for fluid such as pumps at least in one of fluid systems such as a condensate cleanup system are constituted by a material containing metal species such as Zn having an effect of suppressing deposition of radioactivity. Alternatively, the surface of these equipments and pipelines for fluids on which water passes is formed by a coating layer comprising a material containing a metal having a radiation deposition suppressing effect. As a result, radioactivity deposited on the equipments and pipelines for fluids is reduced. In addition, since the method described above may be applied only at least to a portion of the members constituting at least one of the systems for fluids, it is economical. Accordingly, radiation dose upon inspection of equipments and pipelines for fluids can be reduced simply and reliably. (I.S.)

  1. Nuclear power plant

    International Nuclear Information System (INIS)

    Urata, Hidehiro; Oya, Takashi.

    1996-01-01

    The present invention provides a highly safe light water-cooled type nuclear power plant capable of reducing radiation dose by suppressing deposition of activated corrosion products by a simple constitution. Namely, equipments and pipelines for fluid such as pumps at least in one of fluid systems such as a condensate cleanup system are constituted by a material containing metal species such as Zn having an effect of suppressing deposition of radioactivity. Alternatively, the surface of these equipments and pipelines for fluids on which water passes is formed by a coating layer comprising a material containing a metal having a radiation deposition suppressing effect. As a result, radioactivity deposited on the equipments and pipelines for fluids is reduced. In addition, since the method described above may be applied only at least to a portion of the members constituting at least one of the systems for fluids, it is economical. Accordingly, radiation dose upon inspection of equipments and pipelines for fluids can be reduced simply and reliably. (I.S.)

  2. Underground nuclear power plant

    International Nuclear Information System (INIS)

    Takahashi, Hideo.

    1997-01-01

    In an underground-type nuclear power plant, groups of containing cavities comprising a plurality of containing cavities connected in series laterally by way of partition walls are disposed in parallel underground. Controlled communication tunnels for communicating the containing cavities belonging to a control region to each other, and non-controlled communication tunnels for communicating containing cavities belonging to a non-controlled area to each other are disposed underground. A controlled corridor tunnel and a non-controlled corridor tunnel extended so as to surround the containing cavity groups are disposed underground, and the containing cavities belonging to the controlled area are connected to the controlled corridor tunnel respectively, and the containing cavities belonging to the non-controlled area are connected to the non-controlled corridor tunnel respectively. The excavating amount of earth and sand upon construction can be reduced by disposing the containing cavity groups comprising a plurality of containing cavities connected in series laterally. The time and the cost for the construction can be reduced, and various excellent effects can be provided. (N.H.)

  3. 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

  4. 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.)

  5. Development of carbon steel with superior resistance to wall thinning and fracture for nuclear piping system

    International Nuclear Information System (INIS)

    Rhee, Chang Kyu; Lee, Min Ku; Park, Jin Ju

    2010-07-01

    Carbon steel is usually used for piping for secondary coolant system in nuclear power plant because of low cost and good machinability. However, it is generally reported that carbon steel was failed catastrophically because of its low resistance to wall thinning and fracture toughness. Especially, flow accelerated corrosion (FAC) is one of main problems of the wall thinning of piping in the nuclear power plant. Therefore, in this project, fabrication technology of new advanced carbon steel materials modified by dispersion of nano-carbide ceramics into the matrix is developed first in order to improve the resistance to wall thinning and fracture toughness drastically compared to the conventional one. In order to get highly wettable fine TiC ceramic particles into molten metal, the micro-sized TiC particles were first mechanically milled by Fe (MMed TiC/Fe) in a high energy ball mill machine in Ar gas atmosphere, and then mixed with surfactant metal elements (Sn, Cr, Ni) to obtain better wettability, as this lowered surface tension of the carbon steel melt. According to microscopic images revealed that an addition of MMed TiC/Fe-surfactant mixed powders favorably disperses the fine TiC particles in the carbon steel matrix. It was also found that the grain size refinement of the cast matrix is achieved remarkably when fine TiC particles were added due to the fact that they act as nucleation sites during the solidification process. As a results, a cast carbon steel dispersed with fine TiC particles shows improved mechanical properties such as hardness, tensile strength and cavitation resistance compared to that of without particles. However, the slight decrease of toughness was found

  6. Nuclear power plant V-1

    International Nuclear Information System (INIS)

    1998-01-01

    The nuclear power plant Bohunice V -1 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - 24 April 1972; first controlled reactor power - 27 November 1978, 15 March 1980; connection to the grid - 17 December 1978, 26 March 1980; commercial operation - 1 April 1980, 7 January 1981. This leaflet contains: NPP V-1 construction; Major technological equipment (Primary circuit: Nuclear reactor [WWER 440 V230 type reactor];Steam generator; Reactor Coolant Pumps; Primary Circuit Auxiliary Systems. Secondary circuit: Turbine generators, Nuclear power plant electrical equipment; power plant control) and technical data

  7. The 1994 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping

    International Nuclear Information System (INIS)

    Slagis, G.C.

    1995-01-01

    A record of the 1994 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping is provided. The focus of the forum was the design-by-rule method for seismic design of piping. Issues such as acceptance criteria, ductility considerations, demonstration of margin, training, verification and costs were discussed. The use of earthquake experience data, including the recent Northridge earthquake, to justify a design-by-rule method was explored. The majority of the participants felt there are not significant advantages to developing a design-by-rule approach for new plant design. One major disadvantage was considered by many to be training. Extensive training will be required to properly implement a design-by-rule approach. Verification of designs was considered by the majority to be equally important for design-by-rule as for design-by-analysis. If a design-by-rule method is going to be effective, the method will have to be based on ductility considerations (UBC approach). A significant issue will be justification of seismic margins with liberal rules. The UBC approach is being questioned by some because of the recent structural cracking problems in the Northridge earthquake

  8. Battelle integrity of nuclear piping program. Summary of results and implications for codes/standards

    International Nuclear Information System (INIS)

    Miura, Naoki

    2005-01-01

    The BINP(Battelle Integrity of Nuclear Piping) program was proposed by Battelle to elaborate pipe fracture evaluation methods and to improve LBB and in-service flaw evaluation criteria. The program has been conducted from October 1998 to September 2003. In Japan, CRIEPI participated in the program on behalf of electric utilities and fabricators to catch up the technical backgrounds for possible future revision of LBB and in-service flaw evaluation standards and to investigate the issues needed to be reflected to current domestic standards. A series of the results obtained from the program has been well utilized for the new LBB Regulatory Guide Program by USNRC and for proposal of revised in-service flaw evaluation criteria to the ASME Code Committee. The results were assessed whether they had implications for the existing or future domestic standards. As a result, the impact of many of these issues, which were concerned to be adversely affected to LBB approval or allowable flaw sizes in flaw evaluation criteria, was found to be relatively minor under actual plant conditions. At the same time, some issues that needed to be resolved to address advanced and rational standards in the future were specified. (author)

  9. Report of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    2001-12-01

    In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The following findings and conclusion were made as the result of the present examination. (1) Wall thickness of the pipe was significantly reduced in the ruptured region. (2) Dimple pattern resulting from ductile fracture by shearing was observed in the fracture surfaces of nearly all of the pieces and no indication of fatigue crack growth was found. (3) Microstructure showed a typical carbon

  10. Field vibration test of principal equipment of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shiraki, Kazuhiro; Fujita, Katsuhisa; Kajimura, Motohiko; Ikegami, Yasuhiko; Hanzawa, Katsumi; Sakai, Yoshiyuki; Kokubo, Eiji; Igarashi, Shigeru

    1984-09-01

    Japan is one of the most earthquake-stricken countries in the world, and demands for aseismic design have become severer recently. In a nuclear power plant in particular, consisting of a reactor vessel and other facilities dealing with a radioactive substance in some form or other, it is essential from the standpoint of safety to eliminate any possibility of radioactive hazards for the local public, and the employees at the plant as well, if these facilities are struck by an earthquake. This paper is related to the reactor vessel, reactor primary cooling equipment and piping system and important general piping as examples of important facilities of a nuclear power plant, and discusses vibration tests of an actual plant in the field from the standpoint of enhancing the aseismic safety of the Mitsubishi PWR nuclear power plant. Especially concerning vibration test technology, the effects in the evaluation of aseismic safety and its limits are studied to prove how it contributes to the enhancement of the reliability of aseismic design of nuclear power plants.

  11. Man and nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    According to the Inst. fuer Unfallforschung/TUeV Rheinland, Koeln, the interpretation of empirical data gained from the operation of nuclear power plants at home and abroad during the period 1967-1975 has shown that about 38% of all reactor accidents were caused by human failures. These occured either during the design and construction, the commissioning, the reconditioning or the operation of the plants. This very fact stresses human responsibility for the safety of nuclear power plants, in spite of those plants being automated to a high degree and devices. (orig.) [de

  12. Cause Analysis of Flow Accelerated Corrosion and Erosion-Corrosion Cases in Korea Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. S.; Lee, S. H. [Korea Hydro and Nuclear Power Co., Ltd., Gyeongju (Korea, Republic of); Hwang, K. M. [KEPCO Engineering and Construction Company, Gimcheon (Korea, Republic of)

    2016-08-15

    Significant piping wall thinning caused by Flow-Accelerated Corrosion (FAC) and Erosion-Corrosion (EC) continues to occur, even after the Mihama Power Station unit 3 secondary pipe rupture in 2004, in which workers were seriously injured or died. Nuclear power plants in many countries have experienced FAC and EC-related cases in steam cycle piping systems. Korea has also experienced piping wall thinning cases including thinning in the downstream straight pipe of a check valve in a feedwater pump line, the downstream elbow of a control valve in a feedwater flow control line, and failure of the straight pipe downstream of an orifice in an auxiliary steam return line. Cause analyses were performed by reviewing thickness data using Ultrasonic Techniques (UT) and, Scanning Electron Microscope (SEM) images for the failed pipe, and numerical simulation results for FAC and EC cases in Korea Nuclear Power Plants. It was concluded that the main cause of wall thinning for the downstream pipe of a check valve is FAC caused by water vortex flow due to the internal flow shape of a check valve, the main cause of wall thinning for the downstream elbow of a control valve is FAC caused by a thickness difference with the upstream pipe, and the main cause of wall thinning for the downstream pipe of an orifice is FAC and EC caused by liquid droplets and vortex flow. In order to investigate more cases, additional analyses were performed with the review of a lot of thickness data for inspected pipes. The results showed that pipe wall thinning was also affected by the operating condition of upstream equipment. Management of FAC and EC based on these cases will focus on the downstream piping of abnormal or unusual operated equipment.

  13. Radiochemistry in nuclear power plants

    International Nuclear Information System (INIS)

    Schwarz, W.

    2007-01-01

    Radiochemistry is employed in nuclear power plants not as an end in itself but, among other things, as a main prerequisite of optimum radiation protection. Radiochemical monitoring of various loops provides important information about sources of radioactivity, activity distribution in the plant and its changes. In the light of these analytical findings, plant crews are able to take measures having a positive effect on radiation levels in the plant. The example of a BWR plant is used to show, among other things, how radiochemical analyses helped to reduce radiation levels in a plant and, as a consequence, to decrease clearly radiation exposure of the personnel despite higher workloads. (orig.)

  14. Nuclear Power Plant Simulation Game.

    Science.gov (United States)

    Weiss, Fran

    1979-01-01

    Presents a nuclear power plant simulation game which is designed to involve a class of 30 junior or senior high school students. Scientific, ecological, and social issues covered in the game are also presented. (HM)

  15. Technical guidelines for aseismic design of nuclear power plants

    International Nuclear Information System (INIS)

    Park, Y.J.; Hofmayer, C.H.

    1994-06-01

    This document is a translation, in its entirety, of the Japan Electric Association (JEA) publication entitled open-quotes Technical Guidelines for Aseismic Design of Nuclear Power Plants - JEAG 4601-1987.close quotes This guideline describes in detail the aseismic design techniques used in Japan for nuclear power plants. It contains chapters dealing with: (a)the selection of earthquake ground motions for a site, (b) the investigation of foundation and bedrock conditions, (c) the evaluation of ground stability and the effects of ground movement on buried piping and structures, (d) the analysis and design of structures, and (e) the analysis and design of equipment and distribution systems (piping, electrical raceways, instrumentation, tubing and HVAC duct). The guideline also includes appendices which summarize data, information and references related to aseismic design technology

  16. Cost effective snubber reduction program for nuclear power plants

    International Nuclear Information System (INIS)

    Adams, T.M.; Antaki, G.A.; Chang, K.C.

    1985-01-01

    Due to the stringent seismic requirements imposed on nuclear power plants, piping engineers have resorted to the extensive use of snubbers to support nuclear piping systems. The advantage of snubbers is that they provide dynamic restraint while allowing free thermal growth of the pipe. Unfortunately, as more plants go into operation, utilities have to face the costs of strict in-service inspection requirements and risks of unscheduled or extended plant outages associated with snubber failures. The snubber inspection requirements, defined in plant Technical Specifications, require periodic visual inspections of all snubbers and functional tests of a percentage of the plant snubbers, during refueling outages. For a typical 1000 Mw unit this represents from 50 to several hundred snubbers to be functionally tested at each refueling outage. Should failures occur during testing, the sample size must be further increased. Very quickly the costs and risks of extended shutdowns have led the industry to consider, and in many cases implement, snubber reduction programs. At the same time several changes in seismic design criteria have greatly facilitated the reduction of snubbers, making snubber elimination economically and technically attractive. In this paper we examine the costs and benefits of snubber reduction programs and propose a method for evaluating their cost benefits

  17. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-01-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics. (author)

  18. Robotics for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-10-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics.

  19. Decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

    A survey of the main questions of decommissioning of nuclear power plants will be given in the sight of German utilities (VDEW-Working group 'Stillegung'). The main topics are: 1) Definitions of decommissioning, entombment, removal and combinations of such alternatives; 2) Radioactive inventory (build up and decay); 3) Experience up to now; 4) Possibilities to dismantle are given by possibility to repair nuclear power plants; 5) Estimated costs, waste, occupational radiation dose; 6) German concept of decommissioning. (orig./HK) [de

  20. Enhancement of J estimation for typical nuclear pipes with a circumferential surface crack under tensile load

    International Nuclear Information System (INIS)

    Cho, Doo Ho; Woo, Seung Wan; Choi, Jae Boong; Kim, Young Jin; Chang, Yoon Suk; Jhung, Myung Jo; Choi, Young Hwan

    2010-01-01

    This paper is to report enhancement of engineering J estimation for semi-elliptical surface cracks under tensile load. Firstly, limitation of the sole solution suggested by Zahoor is shown for reliable structural integrity assessment of thin-walled nuclear pipes. An improved solution is then developed based on extensive 3D FE analyses employing deformation plasticity theory for typical nuclear piping materials. It takes over the structure of the existing solution but provides new tabulated plastic influence functions to cover a wide range of pipe geometry and crack shape. Furthermore, to facilitate easy prediction of the plastic influence function, an alternative simple equation is also developed by using a statistical response surface method. The proposed H 1 values can be used for elastic-plastic fracture analyses of thin-walled pipes with a circumferential surface crack subjected to tensile loading

  1. Enhancement of J estimation for typical nuclear pipes with a circumferential surface crack under tensile load

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Doo Ho; Woo, Seung Wan; Choi, Jae Boong; Kim, Young Jin [Sungkyunkwan University, Suwon (Korea, Republic of); Chang, Yoon Suk [Kyung Hee University, Yongin (Korea, Republic of); Jhung, Myung Jo; Choi, Young Hwan [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2010-03-15

    This paper is to report enhancement of engineering J estimation for semi-elliptical surface cracks under tensile load. Firstly, limitation of the sole solution suggested by Zahoor is shown for reliable structural integrity assessment of thin-walled nuclear pipes. An improved solution is then developed based on extensive 3D FE analyses employing deformation plasticity theory for typical nuclear piping materials. It takes over the structure of the existing solution but provides new tabulated plastic influence functions to cover a wide range of pipe geometry and crack shape. Furthermore, to facilitate easy prediction of the plastic influence function, an alternative simple equation is also developed by using a statistical response surface method. The proposed H{sub 1} values can be used for elastic-plastic fracture analyses of thin-walled pipes with a circumferential surface crack subjected to tensile loading

  2. A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, Robert A.

    2013-09-18

    Hydrogen explosions may occur simultaneously with water hammer accidents in nuclear facilities, and a theoretical mechanism to relate water hammer to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in pipe systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the pipe system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany. Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism.

  3. Acoustic leak detection in nuclear power plants

    International Nuclear Information System (INIS)

    McElroy, J.W.

    1986-01-01

    For several years now, utilities have been utilizing acoustic leak detection methods as an operating tool in their nuclear power stations. The purpose for using the leak detection system at the various stations vary from safety, ALARA, improved operations, preventive maintenance, or increased plant availability. This paper describes the various acoustic techniques and their application. The techniques are divided into three categories: specific component leakage, intersystem leakage, and pipe through-wall crack leakage. The paper addresses each category in terms of motivation to monitor, method of application and operation, and benefits to be gained. Current requirements are reviewed and analyzed with respect to the acoustic techniques. The paper shows how acoustic leak detection is one of the most effective leak detection tools available. 9 figures, 1 table

  4. Gel structure of the corrosion layer on cladding pipes of nuclear fuels

    Czech Academy of Sciences Publication Activity Database

    Medek, Jiří; Weishauptová, Zuzana

    2009-01-01

    Roč. 393, č. 2 (2009), s. 306-310 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : cladding pipes of nuclear fuels * corrosion layer * zirconium alloys Subject RIV: JF - Nuclear Energetics Impact factor: 1.933, year: 2009

  5. Nuclear Power Plant Mechanical Component Flooding Fragility Experiments Status

    Energy Technology Data Exchange (ETDEWEB)

    Pope, C. L. [Idaho State Univ., Pocatello, ID (United States); Savage, B. [Idaho State Univ., Pocatello, ID (United States); Johnson, B. [Idaho State Univ., Pocatello, ID (United States); Muchmore, C. [Idaho State Univ., Pocatello, ID (United States); Nichols, L. [Idaho State Univ., Pocatello, ID (United States); Roberts, G. [Idaho State Univ., Pocatello, ID (United States); Ryan, E. [Idaho State Univ., Pocatello, ID (United States); Suresh, S. [Idaho State Univ., Pocatello, ID (United States); Tahhan, A. [Idaho State Univ., Pocatello, ID (United States); Tuladhar, R. [Idaho State Univ., Pocatello, ID (United States); Wells, A. [Idaho State Univ., Pocatello, ID (United States); Smith, C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-24

    This report describes progress on Nuclear Power Plant mechanical component flooding fragility experiments and supporting research. The progress includes execution of full scale fragility experiments using hollow-core doors, design of improvements to the Portal Evaluation Tank, equipment procurement and initial installation of PET improvements, designation of experiments exploiting the improved PET capabilities, fragility mathematical model development, Smoothed Particle Hydrodynamic simulations, wave impact simulation device research, and pipe rupture mechanics research.

  6. Ultrasound propagation in steel piping at electric power plant using clamp-on ultrasonic pulse doppler velocity-profile flowmeter

    International Nuclear Information System (INIS)

    Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige

    2008-01-01

    Venturi nozzles are widely used to measure the flow rates of reactor feedwater. This flow rate of nuclear reactor feedwater is an important factor in the operation of nuclear power reactors. Some other types of flowmeters have been proposed to improve measurement accuracy. The ultrasonic pulse Doppler velocity-profile flowmeter is expected to be a candidate method because it can measure the flow profiles across the pipe cross sections. For the accurate estimation of the flow velocity, the incidence angle of ultrasonic entering the fluid should be carefully estimated by the theoretical approach. However, the evaluation of the ultrasound propagation is not straightforward for the several reasons such as temperature gradient in the wedge or mode conversion at the interface between the wedge and pipe. In recent years, the simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation in steel piping and water, using the 3D-FEM simulation code and the Kirchhoff method, as it relates to the flow profile measurements in power plants with the ultrasonic pulse Doppler velocity-profile flowmeter. (author)

  7. Advanced concepts, analysis approaches and criteria for nuclear piping system design

    International Nuclear Information System (INIS)

    Tang, H.T.; Tagart, S.W. Jr.; Tang, Y.K.

    1992-01-01

    Recent research in piping system design and analysis has resulted in advancements on damping values, independent support motion (ISM), static coefficient method, simplified inelastic method and ASME code criteria changes. In the support area, passive type of supports such as energy-absorbing device and gap stopper have been developed. These advancements provide bases for improved and cost-effective design of future nuclear piping systems. (author)

  8. Cleaning device for steam units in a nuclear power plant

    International Nuclear Information System (INIS)

    Sasamuro, Takemi.

    1978-01-01

    Purpose: To prevent radioactive contamination upon dismantling and inspection of steam units such as a turbine to a building containing such units and the peripheral area. Constitution: A steam generator indirectly heated by steam supplied from steam generating source in a separate system containing no radioactivity is provided to produce cleaning steam. A cleaning steam pipe is connected by way of a stop valve between separation valve of a nuclear power plant steam pipe and a high pressure turbine. Upon cleaning, the separation valve is closed, and steam supplied from the cleaning steam pipe is flown into a condenser. The water thus condensated is returned by way of a feed water heater and a condenser to a water storage tank. (Nakamura, S.)

  9. 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)

  10. 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

  11. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  12. Heat pipe effects in nuclear waste isolation: a review

    International Nuclear Information System (INIS)

    Doughty, C.; Pruess, K.

    1985-12-01

    The existence of fractures favors heat pipe development in a geologic repository as does a partially saturated medium. A number of geologic media are being considered as potential repository sites. Tuff is partially saturated and fractured, basalt and granite are saturated and fractured, salt is unfractured and saturated. Thus the most likely conditions for heat pipe formation occur in tuff while the least likely occur in salt. The relative permeability and capillary pressure dependences on saturation are of critical importance for predicting thermohydraulic behavior around a repository. Mineral redistribution in heat pipe systems near high-level waste packages emplaced in partially saturated formations may significantly affect fluid flow and heat transfer processes, and the chemical environment of the packages. We believe that a combined laboratory, field, and theoretical effort will be needed to identify the relevant physical and chemical processes, and the specific parameters applicable to a particular site. 25 refs., 1 fig

  13. Water intaking facility of nuclear power plant

    International Nuclear Information System (INIS)

    Koyama, Kazuhito; Iwata, Nobukatsu; Ochiai, Kanehiro.

    1994-01-01

    In a water intaking facility of a nuclear power plant, a dam is disposed at a position near a sea shore for preventing sea water introduced in open conduit from flowing to the outer sea upon ebbing of tsunamis. The upper end of the dam is set lower than the lower end of a water-intake pipe of a sea water pump of an ordinary system. A water-intake pipe is disposed to such a length that a sea water pump of an emergency system continues to suck the sea water when the water level of the introduced sea water is lowered than the upper end of the dam during the ebb tide. In addition, a means for stopping the operation of the sea water pump of the ordinary system upon starting of the ebb is disposed. Upon reactor scram for occurrence of earthquakes and the like, either the sea water pump in the ordinary system or the seawater pump in the emergency system operates to ensure required amount of sea water for cooling the reactor. In addition, even if the level of the sea water is lowered than the upper end of the dam, since the sea water pump in the emergency system continues to suck sea water, unnecessary suction for sea water by the ordinary sea water pumps can be eliminated. (N.H.)

  14. Data book of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    2002-03-01

    In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The result of the present examination has already been reported to NISA and has also been published as the JAERI-Tech report No.2001-94. This report is a data book containing the detailed data obtained by the present examination. (author)

  15. 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

  16. Piping benchmark problems for the ABB/CE System 80+ Standardized Plant

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1994-07-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the ABB/Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the System 80+ standard design. It will be required that the combined license licensees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System 80+ piping benchmark is a time history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactor is an advanced PWR type

  17. Loviisa nuclear power plant analyzer

    International Nuclear Information System (INIS)

    Porkholm, K.; Nurmilaukas, P.; Tiihonen, O.; Haenninen, M.; Puska, E.

    1992-12-01

    The APROS Simulation Environment has been developed since 1986 by Imatran Voima Oy (IVO) and the Technical Research Centre of Finland (VTT). It provides tools, solution algorithms and process components for use in different simulation systems for design, analysis and training purposes. One of its main nuclear applications is the Loviisa Nuclear Power Plant Analyzer (LPA). The Loviisa Plant Analyzer includes all the important plant components both in the primary and in the secondary circuits. In addition, all the main control systems, the protection system and the high voltage electrical systems are included. (orig.)

  18. Application of fracture mechanics leak-before-break analyses for protection against pipe rupture in SEP plants

    International Nuclear Information System (INIS)

    Copeland, J.F.; Riccardella, P.C.

    1984-01-01

    In accordance with the latest NRC guidance the leak-before-break technique was evaluated for high-energy piping systems in a nuclear power plant. The elements of this evaluation include determination of: 1) largest crack size which will remain stable; 2) leak rate resulting from a crack with length twice the pipe wall thickness; 3) size of crack which will leak at a rate greater than 1 gpm, if 2) results in less than 1 gpm; and 4) analysis of part-through cracks for subcritical crack growth rates to establish in-service inspection (ISI) intervals. Conclusions reached are: 1) The fracture mechanics leak-before-break approach is shown as a viable option to prevent pipe rupture. 2) Austenitic stainless steel pipes possess significant toughness, and large cracks are required for rupture. 3) The net section plastic collapse analysis is more conservative than tearing modulus evaluations. 4) Leak rates are large enough to assure detection well before cracks reach a critical size. 5) In the case studied, subcritical crack growth is slow enough to require ISI intervals of about 10 years to detect part-through cracks

  19. 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.)

  20. TVA's nuclear power plant experience

    International Nuclear Information System (INIS)

    Willis, W.F.

    1979-01-01

    This paper reviews TVA's nuclear power plant design and construction experience in terms of schedule and capital costs. The completed plant in commercial operation at Browns Ferry and six additional plants currently under construction represent the nation's largest single commitment to nuclear power and an ultimate investment of $12 billion by 1986. The presentation is made in three separate phases. Phase one will recapitulate the status of the nuclear power industry in 1966 and set forth the assumptions used for estimating capital costs and projecting project schedules for the first TVA units. Phase two describes what happened to the program in the hectic early 1979's in terms of expansion of scope (particularly for safety features), the dramatic increase in regulatory requirements, vendor problems, stretchout of project schedules, and unprecedented inflation. Phase three addresses the assumptions used today in estimating schedules and plant costs for the next ten-year period

  1. Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

  2. 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.)

  3. On estimation of reliability for pipe lines of heat power plants under cyclic loading

    International Nuclear Information System (INIS)

    Verezemskij, V.G.

    1986-01-01

    One of the possible methods to obtain a quantitative estimate of the reliability for pipe lines of the welded heat power plants under cyclic loading due to heating-cooling and due to vibration is considered. Reliability estimate is carried out for a common case of loading by simultaneous cycles with different amplitudes and loading asymmetry. It is shown that scattering of the breaking number of cycles for the metal of welds may perceptibly decrease reliability of the welded pipe line

  4. 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

  5. Evaluation of LBB margin of nuclear piping systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun [Seoul Nationl Univ., Seoul (Korea, Republic of); Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-04-15

    Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material.

  6. Evaluation of LBB margin of nuclear piping systems

    International Nuclear Information System (INIS)

    Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun; Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok

    1999-04-01

    Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material

  7. ALARA at nuclear power plants

    International Nuclear Information System (INIS)

    Baum, J.W.

    1991-01-01

    Implementation of the ALARA principle at nuclear power plants presents a continuing challenge for health physicists at utility corporate and plant levels, for plant designers, and for regulatory agencies. The relatively large collective doses at some plants are being addressed through a variety of dose reduction techniques. Initiatives by the ICRP, NCRP, NRC, INPO, EPRI, and BNL ALARA Center have all contributed to a heightened interest and emphasis on dose reduction. The NCRP has formed Scientific Committee 46-9 which is developing a report on ALARA at Nuclear Power Plants. It is planned that this report will include material on historical aspects, management, valuation of dose reduction ($/person-Sv), quantitative and qualitative aspects of optimization, design, operational considerations, and training. The status of this work is summarized in this report

  8. World nuclear power plant capacity

    International Nuclear Information System (INIS)

    1991-01-01

    This report provides the background information for statistics and analysis developed by NUKEM in its monthly Market Report on the Nuclear Fuel Cycle. The assessments in this Special Report are based on the continuous review of individual nuclear power plant projects. This Special Report begins with tables summarizing a variety of nuclear power generating capacity statistics for 1990. It continues with a brief review of the year's major events regarding each country's nuclear power program. The standard NUKEM Market Report tables on nuclear plant capacity are given on pages 24 and 25. Owing to space limitations, the first year shown is 1988. Please refer to previous Special Reports for data covering earlier years. Detailed tables for each country list all existing plants as well as those expected by NUKEM to be in commercial operation by the end of 2005. An Appendix containing a list of abbreviations can be found starting on page 56. Only nuclear power plants intended for civilian use are included in this Special Report. Reactor lifetimes are assumed to be 35 years for all light water reactors and 30 years for all other reactor types, unless other data or definite decommissioning dates have been published by the operators. (orig./UA) [de

  9. 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

  10. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    Wood, R.S.

    1979-12-01

    The following list indicates percentage ownership of commercial nuclear power plants by utility companies as of December 1, 1979. The list includes all plants licensed to operate, under construction, docketed for NRC safety and envionmental reviews, or under NRC antitrust review. It does not include those plants announced but not yet under review or those plants formally cancelled. In many cases, ownership may be in the process of changing as a result of antitrust license conditions and hearings, altered financial conditions, changed power needs, and other reasons. However, this list reflects only those ownership percentages of which the NRC has been formally notified

  11. Feasibility of risk-informed technology for japanese nuclear power plants

    International Nuclear Information System (INIS)

    Yoshida, Tomoo; Fujioka, Terutaka; Kirimoto, Yorihiro; Ueda, Nobuyuki; Kinoshita, Izumi; Kashima, Koichi

    2000-01-01

    Risk-informed technology utilizes Probabilistic Safety Assessment for streamlining the maintenance of nuclear power plants. With this technology, plant components are categorized as either high or low-safety-significant components. The Maintenance requirements focuses on high safety-significant components and are relieved for low safety significant ones. This is expected to reduce plant cost while maintaining safety. We investigated especially risk-informed inservice inspection of piping in U.S. nuclear power plants in the interest of determining its feasibility for Japanese plants. Quantitative and qualitative RI-ISI methods were developed by the ASME/Westinghouse Owners Group and EPRI, respectively. These methods have been incorporated in the ASME Section 11 Code Cases and endorsed by the U.S. Nuclear Regulatory Commission. The quantitative method evaluates component segment risks in terms of pipe failure probability calculated with a probabilistic fracture mechanics(PFM) model and pipe failure impact categorization on core damage frequency(CDF) calculated with PSA. The qualitative method uses pipe failure potential categorization derived from the plant service experiences and pipe failure impact on CDF derived from the PSA insight. The PFM model is applicable only to failures from initial welding defects and stress corrosion cracking, therefore it does not cover such significant failure mechanisms found in nuclear power plants as corrosion or high-cycle fatigue, etc.. Thus, a qualitative failure potential categorization method was developed on the basis of the service experiences of the U.S. nuclear power plants, so that appropriate categorization rules must be developed on the service experiences in Japanese plants. Accordingly, we have devised a software framework with a computer-aided system for the selection of risk significant elements. This system consists of a piping failure database module, a piping failure analysis module, and a piping failure potential

  12. Pressure-dependent fragilities for piping components: Pilot study on Davis-Besse Nuclear Power Station

    International Nuclear Information System (INIS)

    Wesley, D.A.; Nakaki, D.K.; Hadidi-Tamjed, H.; Kipp, T.R.

    1990-10-01

    The capacities of four, low-pressure fluid systems to withstand pressures and temperatures above the design levels were established for the Davis-Besse Nuclear Power Station. The results will be used in evaluating the probability of plant damage from Interfacing System Loss of Coolant Accidents (ISLOCA) as part of the probabilistic risk assessment of the Davis-Besse nuclear power station undertaken by EG ampersand G Idaho, Inc. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The probabilities of failure, as a function of internal pressure, are evaluated as well as the variabilities associated with them. Leak rates or leak areas are estimated for the controlling modes of failure. The pressure capacities for the pipes and vessels are evaluated using limit-state analyses for the various failure modes considered. The capacities are dependent on several factors, including the material properties, modeling assumptions, and the postulated failure criteria. The failure modes for gasketed-flange connections, valves, and pumps do not lend themselves to evaluation by conventional structural mechanics techniques and evaluation must rely primarily on the results from ongoing gasket research test programs and available vendor information and test data. 21 refs., 7 figs., 52 tabs

  13. Latina nuclear power plant

    International Nuclear Information System (INIS)

    1976-03-01

    In the period under review, the Latina power plant produced 1009,07 million kWh with a utilization factor of 72% and an availability factor of 80,51%. The disparity between the utilization and availability factors was mainly due to the shutdown of the plant owing to trade union strife. The reasons for non-availability (19,49%) were almost all related to the functioning of the conventional part and the general servicing of the plant (18 September-28 October). During the shutdown for maintenance, an inspection of the steel members and parts of the core stabilizing structure was made in order to check for the familiar oxidation phenomena caused by CO 2 ; the results of the inspection were all satisfactory. Operation of the plant during 1974 was marked by numerous power cutbacks as a result of outages of the steam-raising units (leaks from the manifolds) and main turbines (inspection and repairs to the LP rotors). Since it was first brought into commercial operation, the plant has produced 13,4 thousand million kWh

  14. Maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    Lashgari, Farbod.

    1995-01-01

    This paper is about maintenance of nuclear power plants. In part one, the outage management of nuclear power plants has described. Meaning of the outage and objectives of outage management is given in introduction. The necessity of a long-term outage strategy is shown in chapter one. The main parts of an outage are as follows: Planning; Preparation; Execution, Each of them and also post-outage review have been explained in the followed chapters. Part two deals with technical details of main primary components of nuclear power plant type WWER. After an introduction about WWER reactors, in each chapter first the general and detailed description of main primary components has given and then their maintenance schedules and procedures. Chapter about reactor and steam generator is related to both types of WWER-440 and WWER-1000, but chapter about reactor coolant pump has specified to WWER-1000 to be more in details.(author)

  15. 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.

  16. The Study on Environmental Fatigue Behavior of Low Alloy Steel and Stainless Steel Pipes Using the Simplified Plant Transients

    International Nuclear Information System (INIS)

    Yoo, One; Song, M. S.; Kim, I. Y.; Park, S. H.; Lee, B. S.

    2010-01-01

    Nuclear components categorized as ASME Code Class 1 shall be evaluated for the fatigue and satisfy the fatigue acceptance criteria, CUF(cumulative usage factor) < 1 in accordance with ASME Code. However, recent studies have shown the fatigue evaluation procedure may not give conservative results when the components operate in the water environment. NRC issued Regulatory Guide 1.207 which enforces the new fatigue evaluation method or Fen(environmental fatigue correction factor) method to nuclear plants to be newly constructed. This paper describes the characteristics of the behavior of low alloy and austenitic stainless steel straight pipe related to environmental fatigue, which are obtained by using the method suggested by Regulatory Guide 1.207 and simplified plant transients

  17. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  18. Ductile fracture mechanics methodology for complex cracks in nuclear piping

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.

    1988-02-01

    Limit load and J-integral estimation solutions are developed for circumferentially complex-cracked pipes in bending. The limit load solution is developed using thick-walled cylinder analysis which included the effects of flaw depth accurately. J-integral estimation solutions are developed that are suitable for a wide range of loading from linear elastic, elastic-plastic to net-section yielding of the flawed section. Mode I stress intensity factor solution is developed from experimental compliance data. Two types of J solutions are developed. First, J solutions for determining the J-resistance curve from single load-displacement record are presented. Next, elastic-plastic J solution in the format of EPRI J estimation scheme is presented. The latter solution was used to predict the load carrying capacity of complex-cracked pipes made of Type-304 stainless steel, Inconel 600, and A106 GrB materials. Predictions were compared against pipe tests to demonstrate the accuracy of the limit load and J estimation solutions.

  19. Ductile fracture mechanics methodology for complex cracks in nuclear piping

    International Nuclear Information System (INIS)

    Zahoor, A.

    1988-01-01

    Limit load and J-integral estimation solutions are developed for circumferentially complex-cracked pipes in bending. The limit load solution is developed using thick-walled cylinder analysis which included the effects of flaw depth accurately. J-integral estimation solutions are developed that are suitable for a wide range of loading from linear elastic, elastic-plastic to net-section yielding of the flawed section. Mode I stress intensity factor solution is developed from experimental compliance data. Two types of J solutions are developed. First, J solutions for determining the J-resistance curve from single load-displacement record are presented. Next, elastic-plastic J solution in the format of EPRI J estimation scheme is presented. The latter solution was used to predict the load carrying capacity of complex-cracked pipes made of Type-304 stainless steel, Inconel 600, and A106 GrB materials. Predictions were compared against pipe tests to demonstrate the accuracy of the limit load and J estimation solutions. (orig.)

  20. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Buden, D.; Ranken, W.A.; Koenig, D.R.

    1980-01-01

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  1. Nuclear power plant V-2

    International Nuclear Information System (INIS)

    1998-01-01

    The nuclear power plant Bohunice V -2 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - December 1976; first controlled reactor power - 7 August 1984, 2 August 1985; connection to the grid - 20 August 1984, 9 August 1985; commercial operation - 14 February 1985, 18 December 1985. This leaflet contains: NPP V-2 construction; Major technological equipment [WWER 440 V230 type reactor; Nuclear Power plant operation safety (Safety barriers; Safety systems [Active safety systems, Passive safety systems]); Centralized heat supply system; Scheme of Bohunice V-2 NPP and technical data

  2. Thermal expansion measurement of turbine and main steam piping by using strain gages in power plants

    International Nuclear Information System (INIS)

    Na, Sang Soo; Chung, Jae Won; Bong, Suk Kun; Jun, Dong Ki; Kim, Yun Suk

    2000-01-01

    One of the domestic co-generation plants have undergone excessive vibration problems of turbine attributed to external force for years. The root cause of turbine vibration may be shaft alignment problem which sometimes is changed by thermal expansion and external force, even if turbine technicians perfectly performed it. To evaluate the alignment condition from plant start-up to full load, a strain measurement of turbine and main steam piping subjected to thermal loading is monitored by using strain gages. The strain gages are bonded on both bearing housing adjusting bolts and pipe stoppers which installed in the x-direction of left-side main steam piping near the turbine inlet in order to monitor closely the effect of turbine under thermal deformation of turbine casing and main steam piping during plant full load. Also in situ load of constant support hangers in main steam piping system is measured by strain gages and its results are used to rebalance the hanger rod load. Consequently, the experimental stress analysis by using strain gages turns out to be very useful tool to diagnose the trouble and failures of not only to stationary components but to rotating machinery in power plants

  3. Nuclear power plant

    International Nuclear Information System (INIS)

    Wieser, R.

    1979-01-01

    The reactor pressure vessel consists of two parts. A cylindrical lower part with a hemispherical steel roof is placed at some distance within an equally shaped pressure vessel of concrete. Both vessels are standing on a common bottom plate. The interspace is kept at subpressure. It serves to contain ring galleries, elevator shafts, and power plant components. (GL) [de

  4. Nuclear plant undergrounding

    International Nuclear Information System (INIS)

    Brown, R.C.; Bastidas, C.P.

    1978-01-01

    Under Section 25524.3 of the Public Resources Code, the California Energy Resources Conservation and Development Commission (CERCDC) was directed to study ''the necessity for '' and the effectiveness and economic feasibility of undergrounding and berm containment of nuclear reactors. The author discusses the basis for the study, the Sargent and Lundy (S and L) involvement in the study, and the final conclusions reached by S and L

  5. Challenges for new nuclear plants

    International Nuclear Information System (INIS)

    Bruschi, H.J.

    2000-01-01

    In the past 20 years, numerous new nuclear plant designs have been introduced in the hope of generating a mixture of features and benefits that generated enough enthusiasm amongst the utility industry decision makers to move forward with a new nuclear generation. Not only has there not been enough enthusiasm, there has been little interest in building new plants with advanced features, especially in the U.S. Compounding this predicament are the changing paradigms to which a new plant would be measured. The near hiatus on new plant orders is the clear cause of the significant consolidation in the nuclear industry. Regardless whether the disappearance of old-line nuclear companies is over or not, some paradigms for new generation designs are unmovable, while others are still under discussion as to their role in future plant designs. This paper will address those design goals that Westinghouse deems already having earned the rank of exemplar, and those still open to debate. Because it is my hope that this paper will lead to a fruitful discussion period, I will provide a list of what I feel are the champion design requirements, and those I consider the contenders. (author)

  6. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Nakayama, Ryoichi; Kimura, Motohiko; Abe, Akira

    1993-01-01

    A continuing need exists for automatic or remote-controlled machines or robots which can perform inspection and maintenance tasks in nuclear power plants. Toshiba has developed several types of monofunctional and multi- functional robots for such purposes over the past 20 years, some of which have already been used in actual plants. This paper describes new multifunctional robots for inspection and maintenance. An inspection robot has been applied in an actual plant for two years for performance testing. Maintenance robots for grinding tasks have also been developed, which can be easily teleoperated by the operator using automatic control. These new robots are expected to be applied to actual inspection and maintenance work in nuclear power plants. (author)

  7. Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning

    International Nuclear Information System (INIS)

    Kastner, W.; Erve, M.; Henzel, N.; Stellwag, B.

    1990-01-01

    Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs

  8. Submarine nuclear power plant

    International Nuclear Information System (INIS)

    Enohara, Masami; Araragi, Fujio.

    1980-01-01

    Purpose: To provide a ballast tank, and nuclear power facilities within the containment shell of a pressure resistance structure and a maintenance operator's entrance and a transmission cable cut-off device at the outer part of the containment shell, whereby after the construction, the shell is towed, and installed by self-submerging, and it can be refloated for repairs by its own strength. Constitution: Within a containment shell having a ballast tank and a pressure resisting structure, there are provided nuclear power facilities including a nuclear power generating chamber, a maintenance operator's living room and the like. Furthermore, a maintenance operator's entrance and exit device and a transmission cable cut-off device are provided within the shell, whereby when it is towed to a predetermined a area after the construction, it submerges by its own strength and when any repair inspection is necessary, it can float up by its own strength, and can be towed to a repair dock or the like. (Yoshihara, H.)

  9. Pipe support

    International Nuclear Information System (INIS)

    Pollono, L.P.

    1979-01-01

    A pipe support for high temperature, thin-walled piping runs such as those used in nuclear systems is described. A section of the pipe to be suppported is encircled by a tubular inner member comprised of two walls with an annular space therebetween. Compacted load-bearing thermal insulation is encapsulated within the annular space, and the inner member is clamped to the pipe by a constant clamping force split-ring clamp. The clamp may be connected to pipe hangers which provide desired support for the pipe

  10. Application of 3-dimensional CAD modeling system in nuclear plants

    International Nuclear Information System (INIS)

    Suwa, Minoru; Saito, Shunji; Nobuhiro, Minoru

    1990-01-01

    Until now, the preliminary work for mutual components in nuclear plant were readied by using plastic models. Recently with the development of computer graphic techniques, we can display the components on the graphics terminal, better than with use of plastic model and actual plants. The computer model can be handled, both telescopically and microscopically. A computer technique called 3-dimensional CAD modeling system was used as the preliminary work and design system. Through application of this system, database for nuclear plants was completed in arrangement step. The data can be used for piping design, stress analysis, shop production, testing and site construction, in all steps. In addition, the data can be used for various planning works, even after starting operation of plant. This paper describes the outline of the 3-dimensional CAD modeling system. (author)

  11. Worldwide nuclear-plant performance

    International Nuclear Information System (INIS)

    Surrey, J.; Thomas, S.

    1980-01-01

    The authors compare the performance of different reactor systems to identify the determinants of plant performance, to examine the evidence of technological maturation, and to discover the principal causes of outage or unavailability. In the light of the findings, they discuss the implications for the UK regarding reactor choice and technology development. They make no judgements about the relative merits of nuclear and fossil-fuel plants, or about safety. (author)

  12. Decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Friske, A.; Thiele, D.

    1988-01-01

    The IAEA classification of decommissioning stages is outlined. The international development hitherto observed in decommissioning of nuclear reactors and nuclear power stations is presented. The dismantling, cutting and decontamination methods used in the decommissioning process are mentioned. The radioactive wastes from decommissioning are characterized, the state of the art of their treatment and disposal is given. The radiation burdens and the decommissioning cost in a decommissioning process are estimated. Finally, some evaluation of the trends in the decommissioning process of nuclear power plants is given. 54 refs. (author)

  13. Nuclear power plant diagnostics

    International Nuclear Information System (INIS)

    Hollo, E.; Siklossy, P.

    1982-01-01

    The cooling circuit vibration diagnostic system of the Block 1 of the Paks nuclear power station is described. The automatic online vibration monitoring system consisting presently of 42 acceleration sensors and 9 pressure fluctuation sensors, which could be extended, performs both global and local inspection of the primary cooling circuit and its components. The offline data processing system evaluates the data for failure mode analysis. The software under development will be appropriate for partial preliminary identification of failure reasons during their initial phases. The installation experiences and the preliminary results during the hot operational testing of Block 1 are presented. (Sz.J.)

  14. 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.)

  15. Analysis of nuclear piping system seismic tests with conventional and energy absorbing supports

    International Nuclear Information System (INIS)

    Park, Y.; DeGrassi, G.; Hofmayer, C.; Bezler, P.; Chokshi, N.

    1997-01-01

    Large-scale models of main steam and feedwater piping systems were tested on the shaking table by the Nuclear Power Engineering Cooperation (NUPEC) of Japan, as part of the Seismic Proving Test Program. This paper describes the linear and nonlinear analyses performed by NRC/BNL and compares the results to the test data

  16. Nuclear plant aging research program

    International Nuclear Information System (INIS)

    Eissenberg, D.M.

    1987-01-01

    The U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, has established the Nuclear Plant Aging Research (NPAR) program in its Division of Engineering Technology. Principal contractors for this program include Oak Ridge National Laboratory, Brookhaven National Laboratory, Idaho National Engineering Laboratory, and Pacific Northwest Laboratory. The program goals are: to identify and characterize time-dependent degradation (aging) of nuclear plant safety-related electrical and mechanical components which could lead to loss of safety function; to identify and recommend methods for detecting and trending aging effects prior to loss of safety function so that timely maintenance can be implemented; and to recommend maintenance practices for mitigating the effects of aging. Research activities include prioritization of system and component aging in nuclear plants, characterization of aging degradation of specific components including identification of functional indicators useful for trending degradation, and testing of practical methods and devices for measuring the functional indicators. Aging assessments have been completed on electric motors, snubbers, motor-operated valves, and check valves. Testing of trending methods and devices for motor-operated valves and check valves is in progress

  17. 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)

  18. Nuclear power plant

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1982-01-01

    Purpose: To decrease the reducing speed of nuclear reactor water level after the water level has reached a turbine trip level to trip the turbine thereby preventing cooling systems or the likes from undesired operation upon separation caused by the reduction of the reactor water level to a low water level before the water level control is switched to the manual control. Constitution: Two feedwater pumps arranged in parallel are operated in usual operation to feedwater to a BWR type reactor. If a trouble should occur in a feedwater controller to increase the feedwater rate and the reactor water level, one of the feedwater pumps is tripped by a signal from a feedwater pump trip device. Then, when the trip level is reached again the remaining pump is tripped. In this way, the sudden decrease in the feedwater rate and the reactor water level can be prevented. (Yoshino, Y.)

  19. Operating experience 1993 in Swedish nuclear power plants

    International Nuclear Information System (INIS)

    1995-01-01

    For many years, the Swedish nuclear power plants had a very good track record, compared with the international average. This trend was broken in 1993. During the year, six power plants were shut down for extended periods of time, for different safety-related reasons. During the autumn, a reactor containment leak was detected during scheduled containment leak rate testing at Barsebaeck 2. The unit was shut down for extensive investigation and corrective action for the rest of the year. Ringhals 2 was shut down last six months of the year as crack indications were found in a weld next to a control rod penetration in the reactor vessel head. Extensive tests and analyses revealed that the crack originated from the manufacturing of the vessel head and was of minor importance to safety. Oskarshamn 1 was shut down the whole year. Cracks in cold bent pipes in the residual heat removal system and cracks in the feedwater riser pipes lead to extensive replacement of piping, including pipes inside the reactor vessel. Decontamination of the reactor vessel was successful and attracted world wide interest. A programme for plant status verification was started in order to establish long-term operating conditions. Replacement of the pipe insulation and the inlet strainers in the core and containment spray systems solved the problems with clogging at certain failures in Barsebaeck, Ringhals 1 and Oskarshamn 1 and 2. Six of the reactors had an extremely high availability, of about 90 per cent and more. By year end, eleven of the twelve reactors were in full power operation

  20. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  1. Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

    International Nuclear Information System (INIS)

    Goldberg, A.; Streit, R.D.

    1981-05-01

    Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

  2. High cycle thermal fatigue: benchmark at a Te junction piping system of the nuclear power plant Phenix; Fatigue a grand nombre de cycles: benchmark d'un te de tuyauterie de la centrale Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Gelineau, O.; Simoneau, J.P. [NOVATOME, a Div. of Framatome, 69 - Lyon (France); Roubin, P. [CEA Cadarache, DER, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    This paper presents the studies of the benchmark concerning a high cycle thermal fatigue problem. This benchmark is based on an industrial case, a Te junction piping system of the french FBR Phenix. The main objectives were the comparison of the different methods used by the participants and the analysis of the damage evaluation methods capacity compared to the observed phenomena. This study took place in an international framework with the United Kingdom, Italy, Japan, Korea, Russia, India and France. (A.L.B.)

  3. Passive Nuclear Plants Program (UPDATE)

    International Nuclear Information System (INIS)

    Chimeno, M. A.

    1998-01-01

    The light water passive plants program (PCNP), today Advanced Nuclear Power Plants Program (PCNA), was constituted in order to reach the goals of the Spanish Electrical Sector in the field of advanced nuclear power plants, optimize the efforts of all Spanish initiatives, and increase joint presence in international projects. The last update of this program, featured in revision 5th of the Program Report, reflects the consolidation of the Spanish sector's presence in International programs of the advanced power plants on the basis of the practically concluded American ALWR program. Since the beginning of the program , the PCNP relies on financing from the Electrical sector, Ocide, SEPI-Endesa, Westinghouse, General Electric, as well as from the industrial cooperators, Initec, UTE (Initec- Empresarios Agrupados), Ciemat, Enusa, Ensa and Tecnatom. The program is made up of the following projects, already concluded: - EPRI's Advanced Light Water Plants Certification Project - Westinghouse's AP600 Project - General Electric's SBWR Project (presently paralyzed) and ABWR project Currently, the following project are under development, at different degrees of advance: - EPP project (European Passive Plant) - EBWR project (European Advanced Boiling Water Reactor)

  4. New technical knowledge to be implemented to the revision of rules on pipe wall thinning management for PWR plants

    International Nuclear Information System (INIS)

    Hirai, Junya; Nakamura, Takao; Amano, Yoichi

    2013-01-01

    Rules for PWR plant pipe wall thinning management were formulated by the Japan Society of Mechanical Engineers in 2006. Since then thinning management of Japanese PWR plants has been carried out based on this rule. Pipe wall thinning phenomena to be dealt with in this rule have been identified in many piping components of power plants. New technical knowledge has been accumulated since the issuance of 2006 edition. We have formulated these knowledge and information about the thinning phenomena in PWR power plants. Given the history of application of this rule, we have to make our best effort to carry out a study of latest technical knowledge and implement them to the revision of rule and improve pipe wall thinning management. This paper summarizes the new technical knowledge and basis to be implemented to the revision of rules on pipe wall thinning management for PWR plants in Japan. (author)

  5. Relative costs to nuclear plants: international experience

    International Nuclear Information System (INIS)

    Souza, Jair Albo Marques de

    1992-03-01

    This work approaches the relative costs to nuclear plants in the Brazil. It also presents the calculation methods and its hypothesis to determinate the costs, and the nacional experience in costs of investment, operating and maintenance of the nuclear plants

  6. Cooling water recipients for nuclear power plants

    International Nuclear Information System (INIS)

    Dahl, F.-E.; Saetre, H.J.

    1971-10-01

    The hydrographical and hydrological conditions at 17 prospective nuclear power plant sites in the Oslofjord district are evaluated with respect to their suitability as recipients for thermal discharges from nuclear power plants. No comparative evaluations are made. (JIW)

  7. The 1995 forum on appropriate criteria and methods for seismic design of nuclear piping

    International Nuclear Information System (INIS)

    Slagis, G.C.

    1996-01-01

    A record of the 1995 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping is provided. The focus of the forum was the earthquake experience data base and whether the data base demonstrates that seismic inertia loads will not cause failure in ductile piping systems. This was a follow-up to the 1994 Forum when the use of earthquake experience data, including the recent Northridge earthquake, to justify a design-by-rule method was explored. Two possible topics for the next forum were identified--inspection after an earthquake and design for safe-shutdown earthquake only

  8. Nuclear power plant

    International Nuclear Information System (INIS)

    Inami, Ichiro; Kobayashi, Minoru.

    1995-01-01

    In a condensate cleanup system and a reactor water cleanup system of a BWR-type reactor, in which primary coolants flow, there is disposed a filtering and desalting device using hollow thread membrane filter and ion exchange resin for a condensate cleanup system, and using a high temperature filter made of a metal, a metal oxide or ceramics as a filtering material and a precoat filter made of a powdery ion exchange resin as a filtering material for a reactor water cleanup system. This can completely remove cruds generated in the condensate system. Since the reactor water cleanup system comprises the powdery resin precoat-type filtering and desalting device and the high temperature filter using ceramics, ionic impurities such as radioactive materials can be removed. Accordingly, cruds are not carried into the inside of the reactor, and since the radioactive concentration in the reactor water is reduced, radiation exposure upon periodical inspection can be minimized almost to zero, to attain a clean plant. (T.M.)

  9. Advanced nuclear plant control complex

    International Nuclear Information System (INIS)

    Scarola, K.; Jamison, S.; Manazir, R.M.; Rescorl, R.L.; Harmon, D.L.

    1991-01-01

    An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system which is nuclear qualified for rapid response to changes in plant parameters and a component control system which together provide a discrete monitoring and control capability at a panel in the control room. A separate data processing system, which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs and a large, overhead integrated process status overview board. The discrete indicator and alarm system and the data processing system receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accidental conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof. (author)

  10. Docommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Essmann, J.

    1981-01-01

    The German utilities operating nuclear power plants have long concerned themselves with aspects of decommissioning and for this purpose an engineering company was given a contract to study the entire spectrum of decommissioning. The results of this study have been available in autumn 1980 and it is possible to discuss all the aspects of decommissioning on a new basis. Following these results no change in the design concept of LWR nuclear power plants in operation or under construction is necessary because the techniques, necessary for decommissioning, are fully available today. The technical feasibility of decommissioning for power plants of Biblis A and KRB type has been shown in detail. The calculations of the quantity of waste produced during removal of a nuclear power plant could be confirmed and it could be determined with high procedure. The radiation dose to the decommissioning personnel is in the range of the radiation protection regulations and is in the same range as the radiation dose to the personnel within a yearly inservice inspection. (AF)

  11. Fire prevention in nuclear plants

    International Nuclear Information System (INIS)

    Cayla, J.P.; Jacquet-Francillon, J.; Matarozzo, F.

    2014-01-01

    About 80 fire starts are reported in EDF nuclear power plants every year but only 3 or 4 turn into a real fire and none has, so far, has led to a major safety failure of a nuclear plant. A new regulation has been implemented in july 2014 that strengthens the concept of defense in depth, proposes an approach that is proportionate to the stakes and risks, this proportionality means that the requirements for a power reactor are not the same as for a nuclear laboratory, and imposes an obligation or result rather than of means. The second article deals with the fire that broke out in the waste silo number 130 at La Hague plant in january 1981. The investigation showed that the flammability of the silo content had been underestimated. The third article presents the consequences of the fire that broke out in a power transformer at the Cattenom plant in june 2013. The fire was rapidly brought under control thanks to the immediate triggering of the emergency plan. The article details also the feedback experience of this event. (A.C.)

  12. 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

  13. Development of automatic inspection robot for nuclear power plants

    International Nuclear Information System (INIS)

    Yamada, K.; Suzuki, K.; Saitoh, K.; Sakaki, T.; Ohe, Y.; Mizutani, T.; Segawa, M.; Kubo, K.

    1987-01-01

    This robot system has been developed for automatic inspection of nuclear power plants. The system configuration is composed of vehicle that runs on monorail, the sensors on the vehicle, an image processer that processes the image information from the sensors, a computer that creates the inspection planning of the robot and an operation panel. This system has two main features, the first is the robot control system. The vehicle and the sensors are controlled by the output data calculated in the computer with the three dimensional plant data. The malfunction is recognized by the combination of the results of image processing, information from the microphone and infrared camera. Tests for a prototype automatic inspection robot system have been performed in the simulated main steam piping room of a nuclear power plant

  14. Practical application of computer graphics in nuclear power plant engineering

    International Nuclear Information System (INIS)

    Machiba, Hiroshi; Kawamura, Hirobumi; Sasaki, Norio

    1992-01-01

    A nuclear power plant is composed of a vast amount of equipment, piping, and so on, and six or seven years are required to complete the design and engineering from the initial planning stage to the time of commercial operation. Furthermore, operating plants must be continually maintained and improved for a long period. Computer graphics were first applied to the composite arrangement design of nuclear power plants in the form of 3-dimensional CAD. Subsequently, as the introduction of CAE has progressed, a huge assortment of information has been accumulated in database, and measures have been sought that would permit the convenient utilization of this information. Using computer graphics technologies, improvement of the interface between the user and such databases has recently been accomplished. In response to the growth in environmental consciousness, photo-realistic simulations for artistic design of the interior and overviews showing harmony with the surroundings have been achieved through the application of computer graphics. (author)

  15. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

    Castano, J.; Schivo, M.

    1987-01-01

    A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

  16. Flow accelerated corrosion and life management of the secondary circuit of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Chocron, Mauricio; La Gamma, Ana M.; Fernandez, Narciso; Moyano, Ricardo; Schiersmann, Christian; Ovando, Luis E.; Sainz, Ricardo A.; Keitelman, Alberto

    2003-01-01

    Flow accelerated corrosion is a matter of concern in secondary circuits of nuclear power plants as well as in fossil fired plants. It contributes to the piping wall thinning and to the corrosion products transport to the steam generators. Because it is a generalized corrosion phenomena, could address to extensive failures. In that sense the plants conduct extensive programs of surveillance of piping degradation. Because the problem involves many variables like alloys, water chemistry and hydrodynamics several models have been proposed in the literature. In the present paper the variables have been organized in a spreadsheet which allows the calculation of normalized risk factors. (author)

  17. Practical application of fracture mechanics with consideration of multiaxiality of stress state to degraded nuclear piping

    International Nuclear Information System (INIS)

    Kussmaul, K.; Blind, D.; Herter, K.H.; Eisele, U.; Schuler, X.

    1995-01-01

    Within the scope of a research project nuclear piping components (T-branches and elbows) with dimensions like the primary coolant lines of PWR plants were investigated. In addition to the experimental full scale tests, extensive numerical calculations by means of the finite element method (FEM) as well as fracture mechanics analyses were performed. The applicability of these methods was verified by comparison with the experimental results. The calculation of fracture mechanics parameters as well as the calculated component stress enabled a statement on crack initiation. The failure behavior could be evaluated by means of the multiaxiality of stress state in the ligament (gradient of the quotient of the multiaxiality of stress state q). With respect to practical application on other pressurized components it is shown how to use the procedure (e.g. in a LBB analysis). A quantitative assessment with regard to crack initiation is possible by comparison of the effective crack initiation value J ieff with the calculated component stress. If the multiaxiality of stress state and the q gradient in the ligament of the fracture ligament of the fracture mechanics specimen and the pressurized component to be evaluated is comparable a quantitative assessment is possible as for crack extension and maximum load. If there is no comparability of the gradients a qualitative assessment is possible for the failure behavior

  18. 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

  19. Computer aided piping layout design in radiochemical plants- an improved software package

    International Nuclear Information System (INIS)

    Raju, R.P.; Siddiqui, H.R.

    1995-01-01

    A software package was developed and it was successfully implemented for the piping layout design of the four process cells of the Kalpakkam Reprocessing Project. This paper discusses in detail all the improvements and modifications that are being carried out in the package so that it becomes more meaningful and useful for implementation for the forthcoming radiochemical plants

  20. Fighting fires in nuclear plants

    International Nuclear Information System (INIS)

    Fantom, L.F.; Weldon, G.E.

    1978-01-01

    Since the Browns Ferry incident, the specter of fires at nuclear plants has been the focus of attention by NRC, the utilities, and the public. There are sophisticated hardware and software available - in the form of fire-protection systems and equipment and training and fire-protection programs. Potential fire losses at nuclear faclities can be staggering. Thus, it behooves all those involved to maximize fire-protection security while simultaneously minimizing the chance of human error, which cancels out the effectiveness of the most up-to-date protective systems and devices

  1. Refined analysis of piping systens according to nuclear standard regulations

    International Nuclear Information System (INIS)

    Bisconti, N.; Lazzeri, L.; Strona, P.P.

    1975-01-01

    A number of programs have been selected to perform particular analyses partly coming from available libraries such as SAP 4 for static and dynamic analysis, partly directly written such as TRATE (for thermal analysis), VASTA, VASTB (to perform the analysis required by ASME 3 for pipings of class A and class B), CFRS (for the calculation of floor response spectra etc.). All the programs are automatically linked and directed by a general program (SCATCA for class A and SCATCB for class B pipings). The starting point is a list of the fabrication, thermal, geometrical and seismic data. The geometrical data are plotted (to check for possible errors) and fed to SAP for static and dynamic analysis together with seismic data and thermal data (average temperatures) reelaborated by TRATE 2 code. The raw data from SAP (weight, thermal, fixed points displacements, seismic, other dynamic) are concerned and reordered and fed to COMBIN 2 program together with the other data from thermal analysis (from TRATE 2). From Combin 2 program all the data are listed; each load set to be considered is provided, for each point, with the necessary data (thermal moments, pressure, average temperatures, thermal gradients), all the data from seismic, weight, and other dynamic analysis are also provided. All this amount of data is stored on a file and examined by VASTA code (for class A) or VASTB (for classes B,C) in order to make a decision about the acceptability of the design. Each subprogram may have an independent output in order to check partial results. Details about each program are provided and an exemple is given, together with a discussion of some-particular problems (thermohydraulic set definition, fatigue analysis, etc.)

  2. QA programs in nuclear power plants

    International Nuclear Information System (INIS)

    Ellingson, A.C.

    1976-01-01

    As an overview of quality assurance programs in nuclear power plants, the energy picture as it appears today is reviewed. Nuclear power plants and their operations are described and an attempt is made to place in proper perspective the alleged ''threats'' inherent in nuclear power. Finally, the quality assurance programs being used in the nuclear industry are described

  3. 4. Nuclear power plant component failures

    International Nuclear Information System (INIS)

    1990-01-01

    Nuclear power plant component failures are dealt with in relation to reliability in nuclear power engineering. The topics treated include classification of failures, analysis of their causes and impacts, nuclear power plant failure data acquisition and processing, interdependent failures, and human factor reliability in nuclear power engineering. (P.A.). 8 figs., 7 tabs., 23 refs

  4. Thermal fatigue crack growth in mixing tees nuclear piping - An analytical approach

    International Nuclear Information System (INIS)

    Radu, V.

    2009-01-01

    The assessment of fatigue crack growth due to cyclic thermal loads arising from turbulent mixing presents significant challenges, principally due to the difficulty of establishing the actual loading spectrum. So-called sinusoidal methods represent a simplified approach in which the entire spectrum is replaced by a sine-wave variation of the temperature at the inner pipe surface. The need for multiple calculations in this process has lead to the development of analytical solutions for thermal stresses in a pipe subject to sinusoidal thermal loading, described in previous work performed at JRC IE Petten, The Netherlands, during the author's stage as seconded national expert. Based on these stress distributions solutions, the paper presents a methodology for assessment of thermal fatigue crack growth life in mixing tees nuclear piping. (author)

  5. Water inlet and steam outlet pipes fitted one inside the other for nuclear reactors

    International Nuclear Information System (INIS)

    Mc Donald, B.N.

    1976-01-01

    A description is given of a combined exhaust nozzle and intake pipe system to support a heat exchanger inside a nuclear reactor pressure vessel. It comprises a generally cylindrical part on the exhaust nozzle, the cylindrical part having an inside passage, a flange around the passage and provided with means to secure the exhaust nozzle to the reactor pressure vessel so as to make it fluidtight. The cylindrical part has an aperture inside to take the intake pipe inside the passage so as to enable the intake pipe to project into the heat exchanger. A collar made on the heat exchanger projects from the heat exchanger to the cylindrical nozzle component to establish communication with the inside passage for the fluid [fr

  6. Maintenance of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Quintana, J. M.; Sanchez, J. T.

    2002-01-01

    With this article about the Maintenance in nuclear power plants we will try to give to see the importance of this kind of installations but the problems found by the clients and contractors to face it, and some possible solutions to improve it. It is necessary to understand this problem like something inner to the installation and must be considerate like a benefit for the same. Of course, there must be adequate Sevices Companies in direct relation with the installation that take the responsibility of assuming and understanding the correct fulfillment of the fixed milestones to get the optimal working of the whole plant systems. (Author)

  7. 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.)

  8. Desalting device for nuclear power plants

    International Nuclear Information System (INIS)

    Shinmura, Akira; Mizumachi, Wataru.

    1981-01-01

    Purpose: To prevent heated steams and sea water from mixing even in the failures of heat transfer pipes for sea water heating heat exchangers. Constitution: Heating pipes from the nuclear reactor are kept from direct contact with sea water. Specifically, heat exchange is at first carried out in a first heat exchanger between the heated steams from the nuclear reactor and highly pressurized heat exchanging fluid (For example, water). Then, next heat exchange is carried out in a second heat exchanger between the heat exchanging fluid referred to above and the sea water to increase the temperature of the latter. That is, a closed heat transfer loop under high pressure is provided between the sea water and the heat source on the side of the nuclear reactor, so that the sea water and the heat source of the nuclear reactor are kept from direct contact. (Ikeda, J.)

  9. Nuclear power plant life management

    International Nuclear Information System (INIS)

    Rorive, P.; Berthe, J.; Lafaille, J.P.; Eussen, G.

    1998-01-01

    Several definitions can be given to the design life of a nuclear power plant just as they can be attributed to the design life of an industrial installation: the book-keeping life which is the duration of the provision for depreciation of the plant, the licensed life which corresponds to the duration for which the plant license has been granted and beyond which a new license should be granted by the safety authorities, the design life which corresponds to the duration specified for ageing and fatigue calculations in the design of some selected components during the plant design phase, the technical life which is the duration of effective technical operation and finally the economic life corresponding to the duration of profitable operation of the plant compared with other means of electricity production. Plant life management refers to the measures taken to cope with the combination of licensed, design, technical and economical life. They can include repairs and replacements of components which have arrived to the end of their life due to known degradation processes such as fatigue, embrittlement, corrosion, wear, erosion, thermal ageing. In all cases however, it is of great importance to plan the intervention so as to minimise the economic impact. Predictive maintenance is used together with in-service inspection programs to fulfil this goal. The paper will go over the methodologies adopted in Belgium in all aspects of electrical, mechanical and civil equipment for managing plant life. (author)

  10. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Byrd, J.S.; Fisher, J.J.; DeVries, K.R.; Martin, T.P.

    1987-01-01

    Expert robots enhance a safety and operations in nuclear plants. E.I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligence, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  11. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Byrd, J.S.; Fisher, J.J.; DeVries, K.R.; Martin, T.P.

    1987-01-01

    Expert robots will enhance safety and operations in nuclear plants. E. I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligent, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  12. Modification of the ASME code z-factor for circumferential surface crack in nuclear ferritic pipings

    International Nuclear Information System (INIS)

    Choi, Young Hwan; Chung, Yon Ki; Koh, Wan Young; Lee, Joung Bae

    1996-01-01

    The purpose of this paper is to modify the ASME Code Z-Factor, which is used in the evaluation of circumferential surface crack in nuclear ferritic pipings. The ASME Code Z-Factor is a load multiplier to compensate plastic load with elasto-plastic load. The current ASME Code Z-Factor underestimates pipe maximum load. In this study, the original SC. TNP method is modified first because the original SC. TNP method has a problem that the maximum allowable load predicted from the original SC. TNP method is slightly higher than that measured from the experiment. Then the new Z-Factor is developed using the modified SC. TNP method. The desirability of both the modified SC. TNP method and the new Z-Factor is examined using the experimental results for the circumferential surface crack in pipings. The results show that (1) the modified SC. TNP method is good for predicting the circumferential surface crack behavior in pipings, and (2) the Z-Factor obtained from the modified SC. TNP method well predicts the behavior of circumferential surface crack in ferritic pipings. 30 refs., 13 figs., 4 tabs. (author)

  13. 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

  14. Nuclear power plants and environment

    International Nuclear Information System (INIS)

    Agudo, E.G.; Penteado Filho, A.C.

    1980-01-01

    The question of nuclear power plants is analysed in details. The fundamental principles of reactors are described as well as the problems of safety involved with the reactor operation and the quantity and type of radioactive released to the environment. It shows that the amount of radioactive is very long. The reactor accidents has occurred, as three mile island, are also analysed. (M.I.A.)

  15. Operation of nuclear power plants

    International Nuclear Information System (INIS)

    Severa, P.

    1988-04-01

    The textbook for training nuclear power plant personnel is centred on the most important aspects of operating modes of WWER-440 reactors. Attention is devoted to the steady state operation of the unit, shutdown, overhaul with refuelling, physical and power start-up. Also given are the regulations of shift operation and the duties of individual categories of personnel during the shift and during the change of shifts. (Z.M.). 3 figs., 1 tab

  16. Summary of design of nuclear vessels and piping to ASME III (NB, NC, ND) and vessels to BS 5500

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1992-01-01

    There is a hierarchy of design code requirements for pressurised components, starting with non-nuclear codes as the minimum and progressing through the ASME III nuclear Classes 3, 2, 1. In establishing and assessing the safety justifications of nuclear plants it is important to have an appreciation of the gradation of requirements in the ASME III design rules and how these go beyond non-nuclear component design rules. There are two broad aspects to the structural integrity of pressurised components, namely the achievement of integrity and the demonstration of integrity. The technical requirements of design codes are associated with achieving integrity while the documentary aspects are usually associated with demonstrating integrity. In practice documents also have a part in achieving integrity in the communication of information between different organisations and personnel involved in the design process. It is not possible to assign simple numerical measures to the relative integrity afforded by non-nuclear codes and the three Classes of ASME III. Instead it is necessary to compare the different requirements of the rules for the various technical and documentary aspects. This paper summarises the most important technical and documentary aspects of the three Classes of the ASME III Code for vessels and the non-nuclear code BS 5500. A similar summary is also provided for the three Classes of ASME III rules for piping. The intention is that the paper provides a basis for appreciating the relative integrity afforded by these various rules. (author)

  17. Nuclear power plants - Quality assurance

    International Nuclear Information System (INIS)

    1980-01-01

    This International Standard defines principles for the establishment and implementation of quality assurance programmes during all phases of design, procurement, fabrication, construction, commissioning, operation, maintenance and decommissioning of structures, systems and components of nuclear power plants. These principles apply to activities affecting the quality of items, such as designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, testing, commissioning, operating, inspecting, maintaining, repairing, refuelling and modifying and eventually decommissioning. The manner in which the principles described in this document will be implemented in different organizations involved in a specific nuclear power project will depend on regulatory and contractual requirements, the form of management applied to a nuclear power project, and the nature and scope of the work to be performed by different organizations

  18. Exploiting nuclear plants in time

    International Nuclear Information System (INIS)

    Tran, Lionel

    2011-02-01

    This document outlines that the French fleet of 58 reactors is only 25 year old in average, and that nuclear safety is strongly regulated, and notably relies on improved indicators and on a decennial re-assessment. It outlines that nuclear energy is a response to energy challenges and that it is therefore relevant to operate the nuclear fleet beyond the initially foreseen lifetime (40 years). Due to maintenance and renewal activities, plants are supposed to be safer and more efficient. To guarantee an always safer and more efficient operation in time, five actions are highlighted: decennial controls, installation and equipment modifications, control and anticipation of installation and equipment wear, competencies and ability renewal, better knowledge of techniques and technologies

  19. Community attitudes toward nuclear plants

    International Nuclear Information System (INIS)

    Peelle, E.

    1982-01-01

    Among the many effects of the accident at Three Mile Island are impacts upon other communities that currently host nuclear-power reactors. Because studies on communities' reactions not immediately available, this chapter reviews existing studies and speculates about possible effects. The patterns and variations in impacts on and responses of nuclear host communities have been the subject of studies at Oak Ridge National Laboratory (Oak Ridge, Tennessee) since 1972. This essay presents results from four post-licensing studies of host communities - Plymouth, Massachusetts, and Waterford, Connecticut (PL-1), and Brunswick, North Carolina, and Appling-Toombs counties, Georgia (PL-2) - along with case study and attitude survey information from two additional communities in which reactors are under construction: Hartsville, Tennessee, and Cherokee County, South Carolina. Differences and similarities between the sites have been assessed in terms of differences in input and social structure; factors affecting the generally favorable attitudes toward local nuclear plants are discussed

  20. Nuclear plant analyzer desktop workstation

    International Nuclear Information System (INIS)

    Beelman, R.J.

    1990-01-01

    In 1983 the U.S. Nuclear Regulatory Commission (USNRC) commissioned the Idaho National Engineering Laboratory (INEL) to develop a Nuclear Plant Analyzer (NPA). The NPA was envisioned as a graphical aid to assist reactor safety analysts in comprehending the results of thermal-hydraulic code calculations. The development was to proceed in three distinct phases culminating in a desktop reactor safety workstation. The desktop NPA is now complete. The desktop NPA is a microcomputer based reactor transient simulation, visualization and analysis tool developed at INEL to assist an analyst in evaluating the transient behavior of nuclear power plants by means of graphic displays. The NPA desktop workstation integrates advanced reactor simulation codes with online computer graphics allowing reactor plant transient simulation and graphical presentation of results. The graphics software, written exclusively in ANSI standard C and FORTRAN 77 and implemented over the UNIX/X-windows operating environment, is modular and is designed to interface to the NRC's suite of advanced thermal-hydraulic codes to the extent allowed by that code. Currently, full, interactive, desktop NPA capabilities are realized only with RELAP5

  1. Occupational dose control in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Viktorsson, C.; Lochard, J.; Benedittini, M.; Baum, J.; Khan, T.A.

    1990-01-01

    Reduction in occupational exposure at nuclear power plants is desirable not only in the interest of the health and safety of plant personnel, but also because it enhances the safety and reliability of the plants. This report summarises the current trends of doses to workers at nuclear power plants and the achievements and developments regarding methods for their reduction

  2. Sabotage at Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Purvis, James W.

    1999-07-21

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented.

  3. Nuclear plant life cycle costs

    International Nuclear Information System (INIS)

    Durante, R.W.

    1994-01-01

    Life cycle costs of nuclear power plants in the United States are discussed. The author argues that these costs have been mishandled or neglected. Decommissioning costs have escalated, e.g. from $328 per unit in 1991 to $370 in 1993 for the Sacramento Municipal Utility District, though they still only amount to less than 0.1 cent per kWh. Waste management has been complicated in the U.S. by the decision to abandon civilian reprocessing; by the year 2000, roughly 30 U.S. nuclear power units will have filled their storage pools; dry storage has been delayed, and will be an expense not originally envisaged. Some examples of costs of major component replacement are provided. No single component has caused as much operational disruption and financial penalties as the steam generator. Operation and maintenance costs have increased steadily, and now amount to more than 70% of production costs. A strategic plan by the Nuclear Power Oversight Committee (of U.S. utilities) will ensure that the ability to correctly operate and maintain a nuclear power plant is built into the original design. 6 figs

  4. Sabotage at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Purvis, James W.

    1999-01-01

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented

  5. The maintenance optimization of structural components in nuclear power plants

    International Nuclear Information System (INIS)

    Bryla, P.; Ardorino, F.; Aufort, P.; Jacquot, J.P.; Magne, L.; Pitner, P.; Verite, B.; Villain, B.; Monnier, B.

    1997-10-01

    An optimization process, called 'OMF-Structures', is developed by Electricite de France (EDF) in order to extend the current 'OMF' Reliability Centered Maintenance to piping structural components. The Auxiliary Feedwater System of a 900 MW French nuclear plant has been studied in order to lay the foundations of the method. This paper presents the currently proposed principles of the process. The principles of the OMF-Structures process include 'Risk-Based Inspection' concepts within an RCM process. Two main phases are identified: The purpose of the first phase is to select the risk-significant failure modes and associated elements. This phase consists of two major steps: potential consequences evaluation and reliability performance evaluation. The second phase consists of the definition of preventive maintenance programs for piping elements that are associated with risk-significant failure modes. (author)

  6. Investigation and evaluation of stress-corrosion cracking in piping of light water reactor plants

    International Nuclear Information System (INIS)

    1979-01-01

    In 1975, a Pipe Cracking Study Group, established by the United States Nuclear Regulatory Commission (USNRC), reviewed intergranular stress-corrosion cracking (IGSCC) in Bioling Water Reactors (BWRs) and issued a report. During 1978, IGSCC was reported for the first time in large-diameter piping (> 20 in.) in a BWR in Germany. This discovery, together with the reported questions concerning the interpretation of ultrasonic inspections, led to the activation of a new Pipe Crack Study Group (PCSG) by USNRC. The charter of the new PCSG was expanded: (1) to include review of potential for stress-corrosion cracking in Pressurized Water Reactors (PWRs) as well as BWRs, (2) to examine operating experience in foreign reactors relevant to IGSCC, and (3) to study five specific questions. The PCSG limited the scope of the study to BWR and PWR piping runs and safe ends attached to the reactor pressure vessel. Not considered were components such as the reactor pressure vessel, pumps, valves, steam generators, large steam turbines, etc. Throughout this report, as well as in the title, the safe ends are arbitrarily defined as piping

  7. On the optimization of support positioning and stiffness for piping systems in power plants

    International Nuclear Information System (INIS)

    Collina, A.; Zanaboni, P.; Belloli, M.

    2005-01-01

    The optimal location of supports for the reduction of vibration in piping systems is an interesting structural problem, that can be approached also with the methods used in the updating parameters problems, especially when the cause of the vibration is due to a resonance or a ineffective damping level, and the shift of a critical frequency is an effective remedy. In the proposed paper a frequency domain method of Finite Element model updating is used in order to properly locate the natural frequencies of a given piping system, starting from the nominal condition. The method considers modal parameters, and enables to directly update the physical quantities of the finite element model, i.e., in the considered application, the stiffness and damping of the supporting devices. The model of the structure is made up by beam finite elements, lumped springs and masses, and rigid links. The general procedure is applied here updating the stiffness and location of the supports of the piping according to the comparison among the current frequencies of the piping systems structure and the ones that are required, according to specific requirements. Application to a system similar to a portion of an actual AP1000 (a new Advanced Passive Nuclear Reactor) piping-support layout is presented. (authors)

  8. Failure probability assessment of wall-thinned nuclear pipes using probabilistic fracture mechanics

    International Nuclear Information System (INIS)

    Lee, Sang-Min; Chang, Yoon-Suk; Choi, Jae-Boong; Kim, Young-Jin

    2006-01-01

    The integrity of nuclear piping system has to be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc., are required. Up to now, this has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approaches are considered as an appropriate method for piping system evaluation. The objectives of this paper are to estimate the failure probabilities of wall-thinned pipes in nuclear secondary systems and to propose limited operating conditions under different types of loadings. To do this, a probabilistic assessment program using reliability index and simulation techniques was developed and applied to evaluate failure probabilities of wall-thinned pipes subjected to internal pressure, bending moment and combined loading of them. The sensitivity analysis results as well as prototypal integrity assessment results showed a promising applicability of the probabilistic assessment program, necessity of practical evaluation reflecting combined loading condition and operation considering limited condition

  9. Quality assurance for pipeline installations in nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Approximately 25% of the cost of a nuclear power plant are caused by the pipeline system, i.e. pipes and fixtures. This is an enourmous portion, demanding of the supplier and fitter of these components a very high safety standard. Against the background of the sociopolitical discussion on the safety of nuclear power plants it was a very useful idea of Mannesmann AG, one of the most important manufacturers in this sector, to lay open its quality assurance concept. On the occasion of the symposium ''Safe components for nuclear energy'' held for the expert press it was pointed out that the share of 17% electric energy coming at present from 15 nuclear power plants will increase in future for economic and ecologic reasons. So, it was maintained, the question is not about the ''pro and the con'', but exclusively about the safety standard for nuclear power plants. Specialists from the various branches of the undertaking informed on how this safety concept for the components pipelines and control equipment is realized. (orig./RW) [de

  10. Piping reliability improvement through passive seismic supports

    International Nuclear Information System (INIS)

    Baltus, R.; Rubbers, A.

    1999-01-01

    The nuclear plants designed in the 1970's were equipped with large quantities of snubbers in auxiliary piping systems. The experience revealed a poor performance of snubbers during periodic inspection, while non-nuclear facility piping survived through strong earthquakes. Consequently, seismic design rules evolved towards more realistic criteria and passive dynamic supports were developed to reduce snubber quantities. These solutions improve the pipe reliability during normal operation while reducing the radiation exposure in a sample line is presented with the impact on pipe stresses compared to the results obtained with passive supports named Limit Stops. (author)

  11. Survey of strong motion earthquake effects on thermal power plants in California with emphasis on piping systems. Volume 1, Main report

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-11-01

    Since 1982, there has been a major effort expended to evaluate the susceptibility of nuclear Power plant equipment to failure and significant damage during seismic events. This was done by making use of data on the performance of electrical and mechanical equipment in conventional power plants and other similar industrial facilities during strong motion earthquakes. This report is intended as an extension of the seismic experience data collection effort and a compilation of experience data specific to power plant piping and supports designed and constructed US power piping code requirements which have experienced strong motion earthquakes. Eight damaging (Richter Magnitude 7.7 to 5.5) California earthquakes and their effects on 8 power generating facilities in use natural gas and California were reviewed. All of these facilities were visited and evaluated. Seven fossel-fueled (dual use natural gas and oil) and one nuclear fueled plants consisting of a total of 36 individual boiler or reactor units were investigated. Peak horizontal ground accelerations that either had been recorded on site at these facilities or were considered applicable to these power plants on the basis of nearby recordings ranged between 0.20g and 0.5lg with strong motion durations which varied from 3.5 to 15 seconds. Most US nuclear power plants are designed for a safe shutdown earthquake peak ground acceleration equal to 0.20g or less with strong motion durations which vary from 10 to 15 seconds

  12. Project of mechanical components for nuclear power plants

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do; Farias Brito David, D. de

    1984-01-01

    The equipment foreseen to be part of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design of the components. The design and calculation's concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities are described. (Author) [pt

  13. Design and structural calculation of nuclear power plant mechanical components

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do

    1986-01-01

    The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt

  14. Confirmation of the seismic resistance of nuclear power plant equipment after assembly

    International Nuclear Information System (INIS)

    Kaznovsky, P. S.; Kaznovsky, A. P.; Saakov, E. S.; Ryasnyj, S. I.

    2013-01-01

    It is shown that the natural frequencies and damping decrements of nuclear power plant equipment can only be determined experimentally and directly at the power generation units (reactors) of nuclear power plants under real disassembly conditions for the equipment, piping network, thermal insulation, etc. A computational experimental method is described in which the natural frequencies and damping decrements are determined in the field and the seismic resistance is reevaluated using these values. This method is the basis of the standards document “Methods for confirming the dynamic characteristics of systems and components of the generating units of nuclear power plants which are important for safety” prepared and introduced in 2012.

  15. Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

    Energy Technology Data Exchange (ETDEWEB)

    Mohan, R.; Marshall, C.; Ghadiali, N.; Wilkowski, G. [Battelle, Columbus, OH (United States)

    1997-04-01

    This paper summarizes work on angled through-wall-crack initiation and combined loading effects on ferritic nuclear pipe performed as part of the Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks In Piping an Piping Welds{close_quotes}. The reader is referred to Reference 1 for details of the experiments and analyses conducted as part of this program. The major impetus for this work stemmed from the observation that initially circumferentially oriented cracks in carbon steel pipes exhibited a high tendency to grow at a different angle when the cracked pipes were subjected to bending or bending plus pressure loads. This failure mode was little understood, and the effect of angled crack grown from an initially circumferential crack raised questions about how cracks in a piping system subjected to combined loading with torsional stresses would behave. There were three major efforts undertaken in this study. The first involved a literature review to assess the causes of toughness anisotropy in ferritic pipes and to develop strength and toughness data as a function of angle from the circumferential plane. The second effort was an attempt to develop a screening criterion based on toughness anisotropy and to compare this screening criterion with experimental pipe fracture data. The third and more significant effort involved finite element analyses to examine why cracks grow at an angle and what is the effect of combined loads with torsional stresses on a circumferentially cracked pipe. These three efforts are summarized.

  16. Research needs and improvement of standards for nuclear power plant design

    International Nuclear Information System (INIS)

    Chen, C.; Moreadith, F.L.

    1978-01-01

    The need for research and improvement of code requirements, for both economy and safety reasons is discussed for the following topics relevant to nuclear power plant structural analysis: Earthquake definition; dynamic behavior of reinforced concrete structures under impact loads; design for postulated pipe rupture; code requirements for loading combinations for concrete structures, reinforcing steel splicing, reinforced concrete structural design for thermal effects. (Auth.)

  17. The PEACE PIPE: Recycling nuclear weapons into a TRU storage/shipping container

    International Nuclear Information System (INIS)

    Floyd, D.; Edstrom, C.; Biddle, K.; Orlowski, R.; Geinitz, R.; Keenan, K.; Rivera, M.

    1997-01-01

    This paper describes results of a contract undertaken by the National Conversion Pilot Project (NCPP) at the Rocky Flats Environmental Technology Site (RFETS) to fabricate stainless steel ''pipe'' containers for use in certification testing at Sandia National Lab, Albuquerque to qualify the container for both storage of transuranic (TRU) waste at RFETS and other DOE sites and shipping of the waste to the Waste Isolation Pilot Project (WIPP). The paper includes a description of the nearly ten-fold increase in the amount of contained plutonium enabled by the product design, the preparation and use of former nuclear weapons facilities to fabricate the components, and the rigorous quality assurance and test procedures that were employed. It also describes how stainless steel nuclear weapons components can be converted into these pipe containers, a true ''swords into plowshare'' success story

  18. Estimates of margins in ASME Code strength values for stainless steel nuclear piping

    International Nuclear Information System (INIS)

    Ware, A.G.

    1995-01-01

    The margins in the ASME Code stainless steel allowable stress values that can be attributed to the variations in material strength are evaluated for nuclear piping steels. Best-fit curves were calculated for the material test data that were used to determine allowable stress values for stainless steels in the ASME Code, supplemented by more recent data, to estimate the mean stresses. The mean yield stresses (on which the stainless steel S m values are based) from the test data are about 15 to 20% greater than the ASME Code yield stress values. The ASME Code yield stress values are estimated to approximately coincide with the 97% confidence limit from the test data. The mean and 97% confidence limit values can be used in the probabilistic risk assessments of nuclear piping

  19. 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.)

  20. 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.)

  1. 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.)

  2. Maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    Migaud, D.; Hutin, J.P.; Jouette, I.; Eymond, P.; Devie, P.; Cudelou, C.; Magnier, S.; Frydman, M.

    2016-01-01

    This document gathers different articles concerning the maintenance of the French nuclear power plants. The first article analyses the impact of the recent law on the energetic transition that sets the share of nuclear power at 50% of the electricity produced by 2025. A consequence may be the decommissioning of 17 to 20 reactors by 2025 and the huge maintenance program called 'Grand Carenage' whose aim is to extend operating life over 40 years will have to be re-considered in order to avoid useless expenses. The second article shows that in 2015 the French nuclear reactor fleet got very good results in terms of availability and safety. There were 49 scheduled outages and among them some ended ahead of time. The third article describes the specificities of the maintenance of a nuclear power plant, for instance the redundancy of some systems implies that maintenance has to deal with systems that have never functioned but must be ready to operate at any moment. Another specificity is the complexity of a nuclear power plant that implies an essential phase of preparation for maintenance operations. Because of safety requirements any maintenance operation has to be controlled, checked and may provide feedback. The fourth article presents the 'Grand Carenage' maintenance program that involves the following operations: the replacement of steam generators, the re-tubing of condensers, the replacement of the filtering drums used for cooling water, the testing of the reactor building, the hydraulic test of the primary circuit and the inspection of the reactor vessel. The fifth article focuses on the organization of the work-site for maintenance operations and the example of the Belleville-sur-Loire is described in the sixth article. Important maintenance operations like 'Grand Carenage' requires a strong collaboration with a network of specialized enterprises and as no reactor (except Flamanville EPR) is being built in France, maintenance

  3. Nuclear Plant Integrated Outage Management

    International Nuclear Information System (INIS)

    Gerstberger, C. R.; Coulehan, R. J.; Tench, W. A.

    1992-01-01

    This paper is a discussion of an emerging concept for improving nuclear plant outage performance - integrated outage management. The paper begins with an explanation of what the concept encompasses, including a scope definition of the service and descriptions of the organization structure, various team functions, and vendor/customer relationships. The evolvement of traditional base scope services to the integrated outage concept is addressed and includes discussions on changing customer needs, shared risks, and a partnership approach to outages. Experiences with concept implementation from a single service in 1984 to the current volume of integrated outage management presented in this paper. We at Westinghouse believe that the operators of nuclear power plants will continue to be aggressively challenged in the next decade to improve the operating and financial performance of their units. More and more customers in the U. S. are looking towards integrated outage as the way to meet these challenges of the 1990s, an arrangement that is best implemented through a long-term partnering with a single-source supplier of high quality nuclear and turbine generator outage services. This availability, and other important parameters

  4. Development of advanced RFID application system for nuclear power plant

    International Nuclear Information System (INIS)

    Onda, Kimiharu; Wakabayashi, Eisuke; Arai, Ryota; Shigemi, Ryosuke; Muro, Keiro; Yuda, Shinya

    2008-01-01

    In late years there comes to be close request for traceability of the information such as production control, construction and maintenance record and work history of nuclear power plants. On the other hand, the Radio-frequency identification (RFID) technology that can specify a product and personnel by an electric wave has raised the functionality and versatility as the base technology that can support ubiquitous information society around the mass production industry. In such a background, this article described the developed system, which applied the RFID to nuclear power plants in the areas of production control of the piping manufacture, the construction management and condition monitoring for maintenance works in order to improve their quality and reliability. (T. Tanaka)

  5. Technical strategy map to employing nuclear power plant aging management

    International Nuclear Information System (INIS)

    Sekimura, Naoto; Kanno, Masanori

    2008-01-01

    Stated in this report are back ground of technical strategy map for nuclear power plant aging management, result of the first road map, significance of technical strategy map, introduction scenario, technology map, road map, upgrade in every year, three groups of academia, industry and government, plan of technical strategy map, upgrade system, comprehensive introduction scenario, measures of nuclear power plant aging management in Japan and the world, new inspection system, outline of 'technical strategy map 2008', preparation of technical information bases in industry, academia and government, collaboration of them, safety researches of neutron radiation damage, stress corrosion crack, fatigue, piping thinning, insulation degradation, concrete degradation, thermal aging, evaluation technologies of earthquake resistance, preparation of rules and standards, ideal maintenance, and training talent. (S.Y.)

  6. Simulators for nuclear power plants

    International Nuclear Information System (INIS)

    Ancarani, A.; Zanobetti, D.

    1983-01-01

    The different types of simulator for nuclear power plants depend on the kind of programme and the degree of representation to be achieved, which in turn determines the functions to duplicate. Different degrees correspond to different simulators and hence to different choices in the functions. Training of nuclear power plant operators takes advantage of the contribution of simulators of various degrees of complexity and fidelity. Reduced scope simulators are best for understanding basic phenomena; replica simulators are best used for formal qualification and requalification of personnel, while modular mini simulators of single parts of a plant are best for replay and assessment of malfunctions. Another category consists of simulators for the development of assistance during operation, with the inclusion of disturbance and alarm analysis. The only existing standard on simulators is, at present, the one adopted in the United States. This is too stringent and is never complied with by present simulators. A description of possible advantages of a European standard is therefore offered: it rests on methods of measurement of basic simulator characteristics such as fidelity in values and time. (author)

  7. 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

  8. Nuclear power plant V-2

    International Nuclear Information System (INIS)

    1999-01-01

    In this leaflet the short history of commissioning of Bohunice V-2 NPP is reviewed (beginning of construction December 1976; First controlled reactor power, Reactor Unit 1 (RU1): 7 August 1984, Reactor Unit 2 (RU2): 2 August 1985; Connection to the grid: RU1 20 August 1984, RU2 9 August 1985; Commercial operation: RU1 14 February 1985, RU2 18 December 1985. The scheme of the nuclear reactor WWER 440/V213 is depicted. The major technological equipment are described. Principles of nuclear power plant operation safety (safety barriers, active and passive safety systems, centralized heat supply system, as well as technical data of the Bohunice V-2 NPP are presented

  9. Nuclear power plant V-1

    International Nuclear Information System (INIS)

    1999-01-01

    In this leaflet the short history of commissioning of Bohunice V-1 NPP is reviewed (beginning of construction 24 April 1972; First controlled reactor power, Reactor Unit 1 (RU1): 27 November 1978, Reactor Unit 2 (RU2): 15 March 1980; Connection to the grid: RU1 17 December 1978, RU2 26 March 1980; Commercial operation: RU1 1 April 1980, RU2 7 January 1981. The scheme of the nuclear reactor WWER 440/V230 is depicted. The major technological equipment (primary circuit, nuclear reactor, steam generators, reactor coolant pumps, primary circuit auxiliary systems, secondary circuit, turbine generators, NPP electrical equipment, and power plant control) are described. Technical data of the Bohunice V-1 NPP are presented

  10. Fatigue evaluation of piping systems with limited vibration test data

    International Nuclear Information System (INIS)

    Huang, S.N.

    1990-11-01

    The safety-related piping in a nuclear power plant may be subjected to pump- or fluid-induced vibrations that, in general, affect only local areas of the piping systems. Pump- or fluid-induced vibrations typically are characterized by low levels of amplitudes and a high number of cycles over the lifetime of plant operation. Thus, the resulting fatigue damage to the piping systems could be an important safety concern. In general, tests and/or analyses are used to evaluate and qualify the piping systems. Test data, however, may be limited because of lack of instrumentation in critical piping locations and/or because of difficulty in obtaining data in inaccessible areas. This paper describes and summarizes a method to use limited pipe vibration test data, along with analytical harmonic response results from finite-element analyses, to assess the fatigue damage of nuclear power plant safety-related piping systems. 5 refs., 2 figs., 11 tabs

  11. Slovak Electric, plc, Mochovce Nuclear Power Plant

    International Nuclear Information System (INIS)

    1999-01-01

    In this popular scientific brochure a brief description of construction scheme of Bohunice Nuclear Power Plant is presented. Electricity generation in a nuclear power plant is described. Instrumentation and control system as well as nuclear safety principles applied on the NPP are presented

  12. 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

  13. Development of new Z-factors for the evaluation of the circumferential surface crack in nuclear pipes

    International Nuclear Information System (INIS)

    Choi, Y.H.; Chung, Y.K.; Park, Y.W.; Lee, J.B.

    1997-01-01

    The purpose of this study is to develop new Z-factors to evaluate the behavior of a circumferential surface crack in nuclear pipe. Z-factor is a load multiplier used in the Z-factor method, which is one of the ASME Code Sec. XI's recommendations for the estimation of a surface crack in nuclear pipe. It has been reported that the load carrying capacities predicted from the current ASME Code Z-factors, are not well in agreement with the experimental results for nuclear pipes with a surface crack. In this study, new Z-factors for ferritic base metal, ferritic submerged arc welding (SAW) weld metal, austenitic base metal, and austenitic SAW weld metal are obtained by use of the surface crack for thin pipe (SC.TNP) method based on GE/EPRI method. The desirability of both the SC.TNP method and the new Z-factors is examined using the results from 48 pipe fracture experiments for nuclear pipes with a circumferential surface crack. The results show that the SC.TNP method is good for describing the circumferential surface crack behavior and the new Z-factors are well in agreement with the measured Z-factors for both ferritic and austenitic pipes. (orig.)

  14. Design of nuclear power plants

    International Nuclear Information System (INIS)

    Lobo, C.G.

    1987-01-01

    The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt

  15. 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

  16. 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

  17. Basis for snubber aging research: Nuclear Plant Aging Research Program

    International Nuclear Information System (INIS)

    Brown, D.P.; Palmer, G.R.; Werry, E.V.; Blahnik, D.E.

    1990-01-01

    This report describes a research plan to address the safety concerns of aging in snubbers used on piping and equipment in commercial nuclear power plants. The work is to be performed under Phase 2 of the Snubber Aging Study of the Nuclear Plant Aging Research Program of the US Nuclear Regulatory Commission with the Pacific Northwest Laboratory (PNL) as the prime contractor. Research conducted by PNL under Phase 1 provided an initial assessment of snubber operating experience and was primarily based on a review of licensee event reports. The work proposed is an extension of Phase 1 and includes research at nuclear power plants and in test laboratories. Included is technical background on the design and use of snubbers in commercial nuclear power applications; the primary failure modes of both hydraulic and mechanical snubbers are discussed. The anticipated safety, technical, and regulatory benefits of the work, along with concerns of the NRC and the utilities, are also described. 21 refs., 7 figs., 1 tab

  18. Nuclear power plant operation 2016. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2017-05-15

    A report is given on the operating results achieved in 2016, events important to plant safety, special and relevant repair, and retrofit measures from nuclear power plants in Germany. Reports about nuclear power plants in Belgium, Finland, the Netherlands, Switzerland, and Spain will be published in a further issue.

  19. Environmental survey around EDF nuclear power plants

    International Nuclear Information System (INIS)

    Foulquier, L.

    1992-01-01

    Description of various types of environmental test carried out under the responsibility of the Operator of nuclear power plants in France, with taking Fessenheim nuclear power plant as an example: permanent monitoring of radioactivity, periodic radioecological assessments, main results of measurements taken, showing that there are no detectable effects of the plant on the environment, policy of openness by publication of these results

  20. Inspection indications, stress corrosion cracks and repair of process piping in nuclear materials production reactors

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.; West, S.L.; Nelson, D.Z.

    1991-01-01

    Ultrasonic inspection of Schedule 40 Type 304 stainless steel piping in the process water system of the Savannah River Site reactors has provided indications of discontinuities in less than 10% of the weld heat affected zones. Pipe sections containing significant indications are replaced with Type 304L components. Post removal metallurgical evaluation showed that the indications resulted from stress corrosion cracking in weld heat-affected zones and that the overall weld quality was excellent. The evaluation also revealed weld fusion zone discontinuities such as incomplete penetration, incomplete fusion, inclusions, underfill at weld roots and hot cracks. Service induced extension of these discontinuities was generally not significant although stress corrosion cracking in one weld fusion zone was noted. One set of UT indications was caused by metallurgical discontinuities at the fusion boundary of an extra weld. This extra weld, not apparent on the outer pipe surface, was slightly overlapping and approximately parallel to the weld being inspected. This extra weld was made during a pipe repair, probably associated with initial construction processes. The two nearly parallel welds made accurate assessment of the UT signal difficult. The implications of these observations to the inspection and repair of process water systems of nuclear reactors is discussed

  1. Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

    1986-01-01

    A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

  2. Natural phenomena hazards evaluation of equipment and piping of Gaseous Diffusion Plant Uranium Enrichment Facility

    International Nuclear Information System (INIS)

    Singhal, M.K.; Kincaid, J.H.; Hammond, C.R.; Stockdale, B.I.; Walls, J.C.

    1995-01-01

    In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation

  3. Natural phenomena hazards evaluation of equipment and piping of Gaseous Diffusion Plant Uranium Enrichment Facility

    Energy Technology Data Exchange (ETDEWEB)

    Singhal, M.K.; Kincaid, J.H.; Hammond, C.R.; Stockdale, B.I.; Walls, J.C. [Oak Ridge National Lab., TN (United States). Technical Programs and Services; Brock, W.R.; Denton, D.R. [Lockheed Martin Energy Systems, Inc., Oak Ridge, TN (United States)

    1995-12-31

    In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation.

  4. Commercialization of nuclear power plant decommissioning technology

    International Nuclear Information System (INIS)

    Williams, D.H.

    1983-01-01

    The commercialization of nuclear power plant decommissioning is presented as a step in the commercialization of nuclear energy. Opportunities for technology application advances are identified. Utility planning needs are presented

  5. TOSHIBA CAE system for nuclear power plant

    International Nuclear Information System (INIS)

    Machiba, Hiroshi; Sasaki, Norio

    1990-01-01

    TOSHIBA aims to secure safety, increase reliability and improve efficiency through the engineering for nuclear power plant using Computer Aided Engineering (CAE). TOSHIBA CAE system for nuclear power plant consists of numbers of sub-systems which had been integrated centering around the Nuclear Power Plant Engineering Data Base (PDBMS) and covers all stage of engineering for nuclear power plant from project management, design, manufacturing, construction to operating plant service and preventive maintenance as it were 'Plant Life-Cycle CAE System'. In recent years, TOSHIBA has been devoting to extend the system for integrated intelligent CAE system with state-of-the-art computer technologies such as computer graphics and artificial intelligence. This paper shows the outline of CAE system for nuclear power plant in TOSHIBA. (author)

  6. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1993-06-01

    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 considers significant. Also other events of general interest are reported. The reports also include a summary of the radiation safety of plant personnel and the environment, as well as tabulated data on the plants' production and load factors

  7. Safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Weihe, G. von; Pamme, H.

    2003-01-01

    Experience shows that German nuclear power plants have always been operated reliably and safely. Over the years, the safety level in these plants has been raised considerably so that they can stand any comparison with other countries. This is confirmed by the two reports published by the Federal Ministry for the Environment on the nuclear safety convention. Behind this, there must obviously stand countless appropriate 'good practices' and a safety management system in nuclear power plants. (orig.) [de

  8. Operation of Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1993-03-01

    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 considers safety significant. Also other events of general interest are reported. The reports also include a summary of the radiation safety of plant personnel and the environment, as well as tabulated data on the plants' production and load factors

  9. HVDC transmission from nuclear power plant

    International Nuclear Information System (INIS)

    Yoshida, Yukio; Takenaka, Kiyoshi; Taniguchi, Haruto; Ueda, Kiyotaka

    1980-01-01

    HVDC transmission directly from a nuclear power plant is expected as one of the bulk power transmission systems from distant power generating area. Successively from the analysis of HVDC transmission from BWR-type nuclear power plant, this report discusses dynamic response characteristics of HVDC transmission (double poles, two circuits) from PWR type nuclear power plant due to dc-line faults (DC-1LG, 2LG) and ac-line faults (3LG) near inverter station. (author)

  10. 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.

  11. 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.)

  12. 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.

  13. Using data visualization tools to support degradation assessment in nuclear piping

    International Nuclear Information System (INIS)

    Jyrkama, M.I.; Pandey, M.D.

    2012-01-01

    Nuclear utilities collect a vast amount of in-service inspection data as part of periodic inspection plans and the detailed assessment and monitoring of various degradation mechanisms, such as fretting, corrosion, and creep. In many cases, the focus is primarily on ensuring that the observed minimum or maximum values are within the acceptable regulatory limits, while the rest of the (often costly) surveillance data remains unused and unanalyzed. The objective of this study is to illustrate how data visualization tools can be used effectively to analyze and consider all of the in-service inspection data, and hence provide valuable support for the degradation assessment in nuclear piping. The 2D and 3D visualization tools discussed in this paper were developed mainly in the context of flow accelerated corrosion (FAC) assessment in feeder piping, where the complex pipe geometries and flow conditions have a significant impact on the ultrasonic (UT) wall thickness measurements. The visualization of eddy current inspection results from the assessment of pitting corrosion of steam generator tubing will also be discussed briefly. The visualization tools provide a more comprehensive view of the degree and extent of degradation, and hence directly support the planning of future inspection of critical components by identifying key locations and areas for detailed monitoring. The results furthermore increase the confidence and reliability of fitness-for-service (FFS) assessments and life cycle management (LCM) planning decisions with respect to component repair or replacement. (author)

  14. Basic safety principles for nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Shiguan

    1989-01-01

    To ensure the safety operation of nuclear power plant, one should strictly adhere to the implelmentation of safety codes and the establishment of nuclear safety code system, as well as the applicable basic safety principles of nuclear power plants. This article briefly introduce the importance of nuclear codes and its economic benefits and the implementation of basic safety principles to be accumulated in practice for many years by various countries

  15. Study on flow phenomena at a mixing tee pipe in plants

    International Nuclear Information System (INIS)

    Maeda, Shogo; Kubota, Hiroki; Sugimoto, Katsumi; Takenaka, Nobuyuki; Miyoshi, Koji

    2016-01-01

    Thermal fatigue cracking may initiate at a tee pipe in plants where high and low temperature fluids flow in. The thermal stress fluctuation is caused by the wall temperature fluctuation due to heat transfer of the fluid temperature fluctuation near the wall. In order to elucidate the flow phenomena at a mixing tee pipe to cause temperature fluctuation, a visualization experiment of the flow in mixing section was conducted using a rectangular test section made of acrylic. As a result, the flow pattern was classified by momentum ratio M_R of the main and branch pipes, and it changed from wall jet to deflecting jet on M_R=3.70, and from deflecting jet to impinging jet on M_R=0.64. The jet flow from the branch pipe is swaying at a period of from about 5 s to 10 s. The relationship between the periods of fluctuation and M_R was investigated. The period decreased as M_R increased. (author)

  16. Guidelines and criteria for nuclear piping and support evaluation and design

    International Nuclear Information System (INIS)

    Rehn, D.L.; Stout, D.H. Jr.; Minichiello, J.C.

    1993-05-01

    The EPRI Research Project 2967-2 has set its fundamental goal to be the development of realistic guidelines and criteria for piping and pipe support design and evaluation. The focus is on items that are most critical to utilities and consists of a variety of tasks relating to piping and pipe support design. One objective of this report is to summarize the recommendations from the seven task reports of the first phase of the project and to provide examples of how to use those recommendations. Criteria and methods for evaluating both short and long term system operation are addressed. Benefits gained from applying the recommendations to actual systems are discussed. The report also reviews other work currently being done within the nuclear industry and assesses the impact of that work on the recommended criteria/methods of this project. The second objective of the report is to discuss possible changes needed in the governing codes or licensing commitments in order to implement the recommendations. Finally, the report describes further research which can be done to advance the criteria presented and answer questions concerning applicability of the proposed criteria to designs not tested/investigated. The basic conclusion reached in the project is that many of the criteria/methods used today in piping analysis/design are overly conservative. The report's conclusion is supported by extensive literature searches, tests, and analyses. The report presents a robust source of reference to utilities which wish to implement changes in criteria and methods. Most of the criteria and methodologies described in the seven task reports and summarized in the following sections will require some effort in licensing or Code changes

  17. Qualification of nuclear power plant operations personnel

    International Nuclear Information System (INIS)

    1984-01-01

    With the ultimate aim of reducing the possibility of human error in nuclear power plant operations, the Guidebook discusses the organizational aspects, the staffing requirements, the educational systems and qualifications, the competence requirements, the ways to establish, preserve and verify competence, the specific aspects of personnel management and training for nuclear power plant operations, and finally the particular situations and difficulties to be overcome by utilities starting their first nuclear power plant. An important aspect presented in the Guidebook is the experience in training and qualification of nuclear power plant personnel in various countries: Argentina, Belgium, Canada, Czechoslovakia, France, Federal Republic of Germany, Spain, Sweden, United Kingdom and United States of America

  18. The operation of nuclear power plants

    International Nuclear Information System (INIS)

    Brosche, D.

    1992-01-01

    The duties to be performed in managing the operation of a nuclear power plant are highly diverse, as will be explained in this contribution by the examples of the Grafenrheinfeld Nuclear Power Station. The excellent safety record and the high availabilities of German nuclear power plants demonstrate that their operators have adopted the right approaches. Systematic evaluation of the operating experience accumulated inhouse and in other plants is of great significance in removing weak spots and improving operation. The manifold and complex activities in the structure of organization and of activities in a nuclear power plant require a high degree of division of labor. (orig.) [de

  19. 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

  20. Emergency control centers for nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Guidance is provided for the development and implementation of emergency control centers for nuclear power plants, including nuclear plant control room, nuclear plant company headquarters, emergency control center, and nuclear plant alternate emergency control center. Requirements and recommendations are presented for the mission, communications, instrumentation and equipment associated with each type of control center. Decisional aids, manning requirements and resources are also given; the decision aids cover both the accident assessment and protective action areas. Both normal and alternate means of communications are considered. Off-site emergency control centers, although not covered in the strict sense by this standard, are considered in an appendix