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.)

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

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

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)

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

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.)

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

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

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)

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)

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.

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

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

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)

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.)

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)

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

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.

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

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.

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

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)

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

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

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

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)

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.

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

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

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)

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

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

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

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

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)

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

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)

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

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

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

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.

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

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.)

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.

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

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.

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

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

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.

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)

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)

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

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

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

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

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)

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)

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.

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)

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.

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

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)

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

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

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)

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

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

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)

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

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)

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)

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

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)

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

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.)

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

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)

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

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)

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

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.)

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

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

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

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

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

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)

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)

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

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.

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

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

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)

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

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

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

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

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.)

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.

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

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

Remote controlled in-pipe manipulators for dye-penetrant inspection and grinding of weld roots inside of pipes

International Nuclear Information System (INIS)

Seeberger, E.K.

2000-01-01

Technical plants which have to satisfy stringent safety criteria must be continuously kept in line with the state of art. This applies in particular to nuclear power plants. The quality of piping in nuclear power plants has been improved quite considerably in recent years. By virtue of the very high quality requirements fulfilled in the manufacture of medium-carrying and pressure-retaining piping, one of the focal aspects of in-service inspections is the medium wetted inside of the piping. A remote controlled pipe crawler has been developed to allow to perform dye penetrant testing of weld roots inside piping (ID ≥ 150 mm). The light crawler has been designed such that it can be inserted into the piping via valves (gate valves, check valves,...) with their internals removed. Once in the piping, all crawler movements are remotely controlled (horizontal and vertical pipes incl. the elbows). If indications are found these discontinuities are ground according to a qualified procedure using a special grinding head attached to the crawler with complete extraction of all grinding residues. The in-pipe grinding is a special qualified three (3) step performance that ensures no residual tensile stress (less than 50 N/mm 2 ) in the finish machined austenitic material surface. The in-pipe inspection system, qualified according to both the specifications of the German Nuclear Safety Standards Commission (KTA) and the American Society of Mechanical Engineers (ASME), has already been used successfully in nuclear power plants on many occasions. (author)

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

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

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

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

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)

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.)

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

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.)

Non-metallic structural wrap systems for pipe

International Nuclear Information System (INIS)

Walker, R.H.; Wesley Rowley, C.

2001-01-01

The use of thermoplastics and reinforcing fiber has been a long-term application of non-metallic material for structural applications. With the advent of specialized epoxies and carbon reinforcing fiber, structural strength approaching and surpassing steel has been used in a wide variety of applications, including nuclear power plants. One of those applications is a NSWS for pipe and other structural members. The NSWS is system of integrating epoxies with reinforcing fiber in a wrapped geometrical configuration. This paper specifically addresses the repair of degraded pipe in heat removal systems used in nuclear power plants, which is typically caused by corrosion, erosion, or abrasion. Loss of structural material leads to leaks, which can be arrested by a NSWS for the pipe. The technical aspects of using thermoplastics to structurally improve degraded pipe in nuclear power plants has been addressed in the ASME B and PV Code Case N-589. Using the fundamentals described in that Code Case, this paper shows how this technology can be extended to pipe repair from the outside. This NSWS has already been used extensively in non-nuclear applications and in one nuclear application. The cost to apply this NSWS is typically substantially less than replacing the pipe and may be technically superior to replacing the pipe. (author)

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.

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

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)

Development of life evaluation technology for nuclear power plant components

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-15

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

Nuclear 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.)

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

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

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.

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

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

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.)

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

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)

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)

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

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)

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.

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

Target study of heat supply from Northern Moravia nuclear power plant

International Nuclear Information System (INIS)

Pospisil, V.

The construction is envisaged in Northern Moravia of a nuclear power plant near Blahutovice in the Novy Jicin district. Heat produced by the nuclear power plant will only be used for district heating; process heat will be supplied from local steam sources. An example is discussed of the Prerov locality which currently is supplied from the Prerov heating and power plant (230 MW), a heating plant (36 MW) and from local sources (15 NW). The study estimates that a thermal power of 430 MW will be required at a time of the start of heat supplies from the nuclear power plant. All heat supply pipelines will be designed as a two-tube system divided into sections with section pipe fittings. The number and location of pipe fittings will be selected depending on the terrain configuration. Water of the maximum outlet temperature of 150 degC will be used as a coolant. The control of the system for Northern Moravia is briefly described. (J.P.)

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)

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

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

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

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

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.

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.)

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

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

Accumulation and preparation of nondestructive inspection data for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In recent years, flaws due to stress corrosion cracking (SCC) in stainless steel piping and nickel based alloy welds were detected at nuclear power plants in Japan. Weld Overlay (WOL) has been developed as a repairing method for piping items without removing flaws. Since the inspection techniques for WOL pipes and nickel based alloy welds are not verified enough in Japan, Japan Nuclear Energy Safety Organization (JNES) has been carrying out a six year research project entitled 'Accumulation and Preparation of Nondestructive Inspection Data for Nuclear Power Plants' regarding nondestructive inspection since FY2007. In this research project, detection and sizing capability of SCC by Ultrasonic Testing (UT) are evaluated using mockup tests. In addition, the results of this project and past nondestructive inspection data performed by JNES projects are gathered, and inputted into the database of NDT information. In FY2012, followings were conducted. 1) Analysis for UT measurement results of nickel based alloy weld simulating safe end of reactor vessel outlet nozzle. 2) Analysis for UT measurement results of cast stainless steel piping. 3) Development of interface of UT simulation. 4) Development of nondestructive testing guideline. (author)

Strategies to control zebra mussel fouling at Kewaunee Nuclear Power Plant

International Nuclear Information System (INIS)

Schwartz, D.; Kasper, J.R.; Pisani, W.

1992-01-01

The zebra mussel, Dreissena polymorpha, is currently infesting the Great Lakes. First discovered in Lake St. Clair, it is now widespread in Lakes Erie and Ontario. The initial efforts relating to zebra mussel control at Wisconsin Public Service Corporation's (WPSC) Kewaunee Nuclear Power Plant (KNPP) precipitated from the Nuclear Regulatory Commission's (NRC) Generic Letter 89-13 regarding fouling of service water (SW) systems at nuclear power plants. In the summer of 1990, Stone and Webster Engineering Corporation (Stone and Webster) was contracted to perform an evaluation of known problems within the SW system. The purposes of the study were to evaluate the actual and potential magnitude of these problems, to evaluate corrective actions to resolve the problems, and to prepare recommendations which would adequately address the issues. Two of the recommendations of this study were to continue a zebra mussel monitoring program which WPSC had already implemented and to evaluate various biocide injection programs should one be required for zebra mussel control. The concern of utilities operating power stations which use waters infested with zebra mussels as their source of cooling and/or makeup water is that mussels (both adults and veligers) will enter plant water systems and foul piping and heat exchangers. This type of fouling can restrict flow through piping, process equipment, and heat exchangers. This type of fouling can restrict flow through piping, process equipment, and heat exchangers, thereby increasing head losses and reducing heat transfer capabilities. The greatest concern in that fouling of this type is within safety-related piping and equipment that are components of service water systems at nuclear power plants

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.)

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

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)

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.)

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.)

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.)

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

Public opinion on age-related degradation in nuclear power plants

International Nuclear Information System (INIS)

Matsuda, Toshihiro

2005-01-01

The first objective of this study is to shed light on the public opinion on age-related degradation at nuclear power plants, namely, on how the general public recognizes or views age-related degradation, which is a safety-related issue and one of the factors contributing to accidents and failures which occur at nuclear power plants. The second objective is to look into the impacts of the accident at Mihama Unit 3, which was caused by a failure to check on the piping wall thickness, on the public opinion on age-related degradation. The first survey was conducted in August 2003, followed by the second survey in October 2004, two months after the accident. The surveys found that the age-related degradation is being perceived by people as one of the risk factors that affect the safety of nuclear power plants. The characteristics of the citizens' perceptions toward age-related degradation in the form of piping cracks are that: (a) many respondents feel uneasy but a relatively few people consider that nuclear operators are technologically capable of coping with this problem; (b) many people believe that radioactivity may be released; and (c) numerous respondents consider that signs of cracks must be thoroughly detected through inspections, while on the other hand, a large percentage of the respondents attribute the accident to improper inspections/maintenance. Based on these results, the government and nuclear operators are expected to give most illuminating explanation on the current situation of and remedial measures against age-related degradation at nuclear power plants. As for the effects of the Mihama-3 accident on the public opinion on age-related degradation, it was revealed that the accident has not so significantly affected the general view for the safety of nuclear power plants, but has newly or strongly aroused people's consciousness of two of the risk factors - improper inspections/maintenance and the age-related degradation of piping. (author)

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.

Inspection technology for high pressure pipes

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae H.; Lee, Jae C.; Eum, Heung S.; Choi, Yu R.; Moon, Soon S.; Jang, Jong H

2000-02-01

Various kinds of defects are likely to be occurred in the welds of high pressure pipes in nuclear power plants. Considering the recent accident of Zuruga nuclear power plant in Japan, reasonable policy is strongly requested for the high pressure pipe integrity. In this study, we developed the technologies to inspect pipe welds automatically. After development of scanning robot prototype in the first research year, we developed and implemented the algorithm of automatic tracking of the scanning robot along the weld line of the pipes. We use laser slit beam on weld area and capture the image using digital camera. Through processing of the captures image, we finally determine the weld line automatically. In addition, we investigated a new technology on micro systems for developing micro scanning robotic inspection of the pipe welds. The technology developed in this study is being transferred to the industry. (author)

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

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)

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

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.)

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

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.)

Team collaborative innovation management based on primary pipes automatic welding project

International Nuclear Information System (INIS)

Li Jing; Wang Dong; Zhang Ke

2012-01-01

The welding quality of primary pipe directly affects the safe operation of nuclear power plants. Primary pipe automatic welding, first of its kind in China, is a complex systematic project involving many facets, such as design, manufacturing, material, and on-site construction. A R and D team was formed by China Guangdong Nuclear Power Engineering Co., Ltd. (CNPEC) together with other domestic nuclear power design institutes, and manufacturing and construction enterprises. According to the characteristics of nuclear power plant construction, and adopting team collaborative innovation management mode, through project co-ordination, resources allocation and building production, education and research collaborative innovation platform, CNPEC successfully developed the primary pipe automatic welding technique which has been widely applied to the construction of nuclear power plant, creating considerable economic benefits. (authors)

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

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.)

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

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.

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

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.

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

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.)

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)

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)

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.

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)

Automatic welding machine for piping

International Nuclear Information System (INIS)

Yoshida, Kazuhiro; Koyama, Takaichi; Iizuka, Tomio; Ito, Yoshitoshi; Takami, Katsumi.

1978-01-01

A remotely controlled automatic special welding machine for piping was developed. This machine is utilized for long distance pipe lines, chemical plants, thermal power generating plants and nuclear power plants effectively from the viewpoint of good quality control, reduction of labor and good controllability. The function of this welding machine is to inspect the shape and dimensions of edge preparation before welding work by the sense of touch, to detect the temperature of melt pool, inspect the bead form by the sense of touch, and check the welding state by ITV during welding work, and to grind the bead surface and inspect the weld metal by ultrasonic test automatically after welding work. The construction of this welding system, the main specification of the apparatus, the welding procedure in detail, the electrical source of this welding machine, the cooling system, the structure and handling of guide ring, the central control system and the operating characteristics are explained. The working procedure and the effect by using this welding machine, and the application to nuclear power plants and the other industrial field are outlined. The HIDIC 08 is used as the controlling computer. This welding machine is useful for welding SUS piping as well as carbon steel piping. (Nakai, Y.)

Wall thinning trend analyses for secondary side piping of Korean NPPs

International Nuclear Information System (INIS)

Hwang, K.M.; Jin, T.E.; Lee, S.H.; Jeon, S.C.

2003-01-01

Since the mid-1990s, nuclear power plants in Korea have experienced wall thinning, leaks, and ruptures of secondary side piping caused by flow-accelerated corrosion (FAC). The pipe failures have increased as operating time progresses. In order to prevent the FAC-induced pipe failures and to develop an effective FAC management strategy, KEPRI and KOPEC have conducted a study for developing systematic FAC management technology for secondary side piping of all Korean nuclear power plants. As a part of the study, FAC analyses were performed using the CHECWORKS code. The analysis results were used to select components for inspection and to determine inspection intervals on each nuclear power plant. This paper describes the introduction of the FAC analysis method and the wall thinning trend analysis results by reactor type, system, and water treatment amine. This paper also represents the site application feasibility for secondary side piping management. The site application feasibility of the analysis results was proven by comparisons of predicted and measured wear rates. (author)

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

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

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.

Application of numerical analysis technique to make up for pipe wall thinning prediction program

International Nuclear Information System (INIS)

Hwang, Kyeong Mo; Jin, Tae Eun; Park, Won; Oh, Dong Hoon

2009-01-01

Flow Accelerated Corrosion (FAC) leads to wall thinning of steel piping exposed to flowing water or wet steam. Experience has shown that FAC damage to piping at fossil and nuclear plants can lead to costly outages and repairs and can affect plant reliability and safety. CHEWORKS have been utilized in domestic nuclear plants as a predictive tool to assist FAC engineers in planning inspections and evaluating the inspection data to prevent piping failures caused by FAC. However, CHECWORKS may be occasionally left out local susceptible portions owing to predicting FAC damage by pipeline group after constructing a database for all secondary side piping in nuclear plants. This paper describes the methodologies that can complement CHECWORKS and the verifications of the CHECWORKS prediction results in terms of numerical analysis. FAC susceptible locations based on CHECWORKS for the two pipeline groups of a nuclear plant was compared with those of numerical analysis based on FLUENT.

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.)

Development of Wall-Thinning Evaluation Procedure for Nuclear Power Plant Piping—Part 1: Quantification of Thickness Measurement Deviation

Directory of Open Access Journals (Sweden)

Hun Yun

2016-06-01

Full Text Available Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs. Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

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)

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)

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

Design considerations for the protection from the effects of pipe rupture

International Nuclear Information System (INIS)

1975-11-01

The methods which are employed by Ebasco Services, Inc. to satisfy the requirement of 10 CFR 50, Appendix A, General Design Criterion (GDC) 4 are discussed. This criterion provides the design basis for protection against the dynamic effects of postulated piping failures (ruptures and cracks) in Nuclear Power Plants. The criteria for postulating pipe failure locations, determining the dynamic effects associated with the postulated failure and designing the plant to satisfactorily withstand postulated pipe failures have been the subject of a great deal of recent work by the nuclear power industry and the Nuclear Regulatory Commission (NRC). This topical report is largely based upon that work as well as upon Ebasco's independent development of analytical tools to aid in the plant design process

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

Integrated piping structural analysis system

International Nuclear Information System (INIS)

Motoi, Toshio; Yamadera, Masao; Horino, Satoshi; Idehata, Takamasa

1979-01-01

Structural analysis of the piping system for nuclear power plants has become larger in scale and in quantity. In addition, higher quality analysis is regarded as of major importance nowadays from the point of view of nuclear plant safety. In order to fulfill to the above requirements, an integrated piping structural analysis system (ISAP-II) has been developed. Basic philosophy of this system is as follows: 1. To apply the date base system. All information is concentrated. 2. To minimize the manual process in analysis, evaluation and documentation. Especially to apply the graphic system as much as possible. On the basis of the above philosophy four subsystems were made. 1. Data control subsystem. 2. Analysis subsystem. 3. Plotting subsystem. 4. Report subsystem. Function of the data control subsystem is to control all information of the data base. Piping structural analysis can be performed by using the analysis subsystem. Isometric piping drawing and mode shape, etc. can be plotted by using the plotting subsystem. Total analysis report can be made without the manual process through the reporting subsystem. (author)

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)

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

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.)

High-cycle fatigue properties of small-bore socket-welded pipe joint

International Nuclear Information System (INIS)

Maekawa, Akira; Noda, Michiyasu; Suzuki, Michiaki

2009-01-01

Piping and equipment in nuclear power plants are structures including many welded joints. Reliability of welded joints is one of high-priority issues to improve the safety of nuclear power plants. However, occurrence of fatigue failures in small-bore socket-welded pipe joints by high-cycle vibrations is still reported. In this study, fatigue experiments on a socket-welded joint of austenitic stainless steel pipe was conducted under excitation conditions similar to those in actual plants to investigate vibration characteristics and fatigue strength. It was found that the natural frequency of pipe with socket-welded joint gradually decreased as fatigue damage developed, according to the Miner rule for fatigue life evaluation. The results indicate that the fatigue life of the welded pipe joint could be estimated by monitoring the decreasing ratio of the natural frequency of the pipe. The evaluation of decreasing ratio of the natural frequency in addition to fatigue damage evaluation by the Miner rule could enhance the accuracy of fatigue life evaluation. (author)

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)

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

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)

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

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

New assessment of feed water piping in GKN I including optimisation of piping supports

International Nuclear Information System (INIS)

Zaiss, W.; Heil, C.; Baier, B.; Manke, A.

2003-01-01

The quality of nuclear power plant components and piping is specified according to the then current state of knowledge. In operation, the quality can be reduced by ageing phenomena, so in-service quality assessment is constantly required. The contribution discusses the individual aspects of reassessment and its technical procedure, using the example of a feedwater pipe in the GKN I containment. (orig.) [de

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.

Replacement of major nuclear power plant components for service life extension

International Nuclear Information System (INIS)

Novak, S.

1987-01-01

Problems are discussed associated with replacement of nuclear power plant components with the aim to extend their original scheduled life. The existing foreign experience shows that it is technically feasible to replace practically all basic components for which the necessity of replacement is established. Data is summed up on the replacement of steam generators in US and West German nuclear power plants showing the duration of the job, the total consumption of manhours, the collective dose equivalent and the cost. Attention is also focused on implemented and projected replacements of circulation pipes in nuclear power plants abroad. Based on these figures, the cost is estimated of the replacement of the reactor vessel and the steam generators for WWER-440 nuclear power plants. The conclusion is arrived at that even based on a conservative estimate, the extension by 20 years of the service life of a nuclear power plant is economically more effective than the construction of a new plant. (Z.M.) 2 tabs., 15 refs., 3 figs

Development of a screening procedure for vibrational fatigue in small bore piping

International Nuclear Information System (INIS)

Smith, J.K.; Riccardella, P.C.; Gosselin, S.R.

1995-01-01

Approximately 80% of the documented fatigue failures in nuclear power plants are caused by high cycle vibrational fatigue. These failures typically occur in socket welded pipe fittings in small bore piping (2 in. nominal diameter and smaller). These failures have been unexpected, and have caused costly, unscheduled outages in some cases. In order to reduce the number of vibrational fatigue failures in operating nuclear power plants, a vibrational fatigue screening procedure has been developed under Electric Power Research Institute (EPRI) sponsorship. The purpose of this paper is to describe this procedure, and to discuss topics related to vibrational fatigue failures. These topics include sources of vibration in nuclear power plants, the effect of socket welds on vibrational fatigue failures, vibrational fatigue screening criteria for small bore piping systems, and good design practices for reducing the number of vibrational fatigue failures in small bore piping

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)

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

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.

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)

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

Base-plate effects on pipe-support stiffness

International Nuclear Information System (INIS)

Winkel, B.V.; LaSalle, F.R.

1981-01-01

Present nuclear power plant design methods require that pipe support spring rates be considered in the seismic design of piping systems. Base plate flexibility can have a significant effect on the spring rates of these support structures. This paper describes the field inspection, test, and analytical techniques used to identify and correct excessively flexible base plates on the Fast Flux Test Facility pipe support structures

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

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

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.

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

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.

Earthquake free design of pipe lines

International Nuclear Information System (INIS)

Kurihara, Chizuko; Sakurai, Akio

1974-01-01

Long structures such as cooling sea water pipe lines of nuclear power plants have a wide range of extent along the ground surface, and are incurred by not only the inertia forces but also forces due to ground deformations or the seismic wave propagation during earthquakes. Since previous reports indicated the earthquake free design of underground pipe lines, it is discussed in this report on behaviors of pipe lines on the ground during earthquakes and is proposed the aseismic design of pipe lines considering the effects of both inertia forces and ground deformations. (author)

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.

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.

Development of a Short-term Failure Assessment of High Density Polyethylene Pipe Welds - Application of the Limit Load Analysis -

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho-Wan; Han, Jae-Jun; Kim, Yun-Jae [Korea University, Seoul (Korea, Republic of); Kim, Jong-Sung [Sunchon National University, Suncheon (Korea, Republic of); Kim, Jeong-Hyeon; Jang, Chang-Heui [KAIST, Daejeon (Korea, Republic of)

2015-04-15

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.

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

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

Full-scale impact test data for tornado-missile design of nuclear plants

International Nuclear Information System (INIS)

Stephenson, A.E.; Sliter, G.E.

1977-01-01

It is standard practice to consider the effects of low-probability impacts of tornado-borne debris (''tornado missiles'' such as utility poles and steel pipes) in the structural design of nuclear power plants in the United States. To provide data that can be used directly in the design procedure, a series of full-scale tornado-missile impact tests was performed. This paper is a brief summary of the results and conclusions from these tests. The tests consisted of reinforced concrete panels impacted by poles, pipes, and rods propelled by a rocket sled. The panels were constructed to current minimum standards and had thicknesses typical of auxiliary buildings of nuclear power plants. A specific objective was the determination of the impact velocities below which the panels do not experience backface scabbing. Another objective was to assess the adequacy of (1) conventional design formulae for penetration and scabbing and (2) conventional design methods for overall structural response. Test missiles and velocities represented those in current design standards. Missiles included utility poles, steel pipes, and steel bars. It is important to interpret the data in this paper in recognition that the test conditions represent conservative assumptions regarding maximum wind speeds, injection of the missile into the wind stream, aerodynamic trajectory, and orientation of missile at impact. Even with the severe assumptions made, the full-scale tests described demonstrate the ability of prototypical nuclear plant walls and roofs to provide adequate protection against postulated tornado-missile impact

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

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.

Pipe damping studies

International Nuclear Information System (INIS)

Ware, A.G.

1986-01-01

The Idaho National Engineering Laboratory (INEL) is conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for use in the dynamic analysis of nuclear power plant piping systems. This paper describes four tasks in the program that were undertaken in FY-86. In the first task, tests were conducted on a 5-in. INEL laboratory piping system and data were analyzed from a 6-in. laboratory system at the ANCO Engineers facility to investigate the parameters influencing damping in the seismic frequency range. Further tests were conducted on 3- and 5-in. INEL laboratory piping systems as the second task to determine damping values representative of vibrations in the 33 to 100 Hz range, typical of hydrodynamic transients. In the third task a statistical evaluation of the available damping data was conduted to determine probability distributions suitable for use in probabilistic risk assessments (PRAs), and the final task evaluated damping data at high strain levels

Defect Depth Measurement of Straight Pipe Specimen Using Shearography

International Nuclear Information System (INIS)

Chang, Ho Seob; Kim, Kyung Suk

2012-01-01

In the nuclear industry, wall thinning defect of straight pipe occur the enormous loss in life evaluation and safety evaluation. To use non-destructive technique, we measure deformation, vibration, defect evaluation. But, this techniques are a weak that is the measurement of the wide area is difficult and the time is caught long. In the secondary side of nuclear power plants mostly used steel pipe, artificiality wall thinning defect make in the side and different thickness make to the each other, wall thinning defect part of deformation measure by using shearography. In addition, optical measurement through deformation, vibration, defect evaluation evaluate pipe and thickness defects of pressure vessel is to evaluate quantitatively. By shearography interferometry to measure the pipe's internal wall thinning defect and the variation of pressure use the proposed technique, the quantitative defect is to evaluate the thickness of the surplus. The amount of deformation use thickness of surplus prediction of the actual thickness defect and approximately 7 percent error by ensure reliability. According to pressure the amount of deformation and the thickness of the surplus through DB construction, nuclear power plant pipe use wall thinning part soundness evaluation. In this study, pressure vessel of thickness defect measure proposed nuclear pipe of wall thinning defect prediction and integrity assessment technology development. As a basic research defected theory and experiment, pressure vessel of advanced stability and soundness and maintainability is expected to contribute foundation establishment

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.

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

Developments of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

Energy Technology Data Exchange (ETDEWEB)

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

2003-03-15

The objective of this research is to develop an efficient evaluation technology and to investigate applicability of newly-developed technology, such as internet-based cyber platform, to operating power plants. Development of efficient evaluation systems for Nuclear Power Plant components, based on structural integrity assessment techniques, are increasingly demanded for safe operation with the increasing operating period of Nuclear Power Plants. The following five topics are covered in this project: development of assessment method for wall-thinned nuclear piping based on pipe test; development of structural integrity program for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for mam components of NPP; development of internet-based cyber platform and integrity program for primary components of NPP; effect of aging on strength of dissimilar welds.

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.

Environmental Assisted Fatigue Evaluation of Direct Vessel Injection Piping Considering Thermal Stratification

International Nuclear Information System (INIS)

Kim, Taesoon; Lee, Dohwan

2016-01-01

As the environmentally assisted fatigue (EAF) due to the primary water conditions is to be a critical issue, the fatigue evaluation for the components and pipes exposed to light water reactor coolant conditions has become increasingly important. Therefore, many studies to evaluate the fatigue life of the components and pipes in LWR coolant environments on fatigue life of materials have been conducted. Among many components and pipes of nuclear power plants, the direct vessel injection piping is known to one of the most vulnerable pipe systems because of thermal stratification occurred in that systems. Thermal stratification occurs because the density of water changes significantly with temperature. In this study, fatigue analysis for DVI piping using finite element analysis has been conducted and those results showed that the results met design conditions related with the environmental fatigue evaluation of safety class 1 pipes in nuclear power plants. Structural and fatigue integrity for the DVI piping system that thermal stratification occurred during the plant operation has conducted. First of all, thermal distribution of the piping system is calculated by computational fluid dynamic analysis to analyze the structural integrity of that piping system. And the fatigue life evaluation considering environmental effects was carried out. Our results showed that the DVI piping system had enough structural integrity and fatigue life during the design lifetime of 60 years

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)

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)

Seismic Capacity Estimation of Steel Piping Elbow under Low-cycle Fatigue Loading

Energy Technology Data Exchange (ETDEWEB)

Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of); Hahm, Dae Gi [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In some cases, this large relative displacement can increase seismic risk of the isolated facility. Especially, a inelastic behavior of crossover piping system to connect base isolated building and fixed base building can caused by a large relative displacement. Therefore, seismic capacity estimation for isolated piping system is needed to increase safety of nuclear power plant under seismic condition. Dynamic behavior analysis of piping system under seismic condition using shake table tests was performed by Touboul et al in 1995. In accordance with their study, plastic behavior could be occurred at pipe elbow under seismic condition. Experimental researches for dynamic behavior of typical piping system in nuclear power plant have been performed for several years by JNES(Japan Nuclear Energy Safety Organization) and NUPEC(Nuclear Power Engineering Corporation). A low cycle ratcheting fatigue test was performed with scaled model of elbow which is a weakest component in piping system by Mizuno et al. In-plane cyclic loading tests under internal pressure condition were performed to evaluate the seismic capacity of the steel piping elbow. Leakage phenomenon occurred on and near the crown in piping elbow. Those cracks grew up in axial direction. The fatigue curve was estimated from test results. In the fatigue curve, loading amplitude exponentially decreased as the number of cycles increased. A FEM model of piping elbow was modified with test results. The relationships between displacement and force from tests and numerical analysis was well matched.

Study on concept of web-based reactor piping design data platform

International Nuclear Information System (INIS)

Wang Yu; Zhou Yu; Dong Jianling; Meng Yang

2005-01-01

For solving the piping design problems such as design data deficiency, designer communication inconvenience and design project inconsistence, Reactor Piping Design Database Platform, which is the main part of the Integrated Nuclear Project Research Platform, is proposed by analyzing the nuclear piping designs in detail. The functions and system structures of the platform are described in the paper for the sake of the realization of the Reactor Piping Design Database Platform. The platform is constituted by web-based management interface, AutoPlant selected as CAD software, and relation database management system (DBMS). (authors)

Development of wall thinning screening system and its application to a commercial nuclear power plant

International Nuclear Information System (INIS)

Ryu, Kyung Ha; Hwang, Il Soon; Kim, Ji Hyun

2013-01-01

Highlights: • Wall loss screening system (WalSS) has been developed based on ES-DCPD method. • Screening criteria was established based on the thinning of the actual shape that occur in the power plant. • With the criteria, the WalSS gives priority of the need for inspection. • This technique was successfully applied to commercial nuclear power plant. - Abstract: A new non-destructive evaluation (NDE) method has been developed for metal pipes for the detection wall thinning. The method has been showed to be suitable for applications to electric power generation plants where flow accelerated corrosion (FAC) of carbon steel piping is a significant cause of increased maintenance and plant personnel casualty. The wall thinning screening system (WalSS) was developed in two major phases. In the first phase, the equipotential switching direct current potential drop (ES-DCPD) method was developed for piping wall (Ryu et al., 2008a, 2010). In the second phase, in this paper, a quantitative detection criteria was developed. The relative ES-DCPD change of 3.8% has been defined as the screening criteria for wall thinning schematization. This criteria means that the component with measured ES-DCPD change greater than 3.8% is called for a more comprehensive examination. In the criteria development, all variables were taken into consideration based on commercial plant piping inspection data such as initial thickness distributions, wall thinning shape and nominal thickness. The developed WalSS based on ES-DCPD was applied to a moisture separator reheater (MSR) drain line of a commercial nuclear power plant (NPP) during a scheduled overhaul. The measured ES-DCPD change was 2.16%, which is lower than the ES-DCPD criteria, identifying the pipe having adequate wall thickness. This is confirmed by site thickness inspection using ultrasonic technique (UT)

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)

Evaluation methods of vibration stress of small bore piping

Energy Technology Data Exchange (ETDEWEB)

Hiramatsu, Miki; Sasaki, Toru [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Fatigue fracture by vibration stress is one of the main causes of troubles which occur at small bore piping in nuclear power plants. Therefore at the plants they manage small bore piping using a method in which their vibration accelerations are measured and the vibration stresses are calculated. In this work, vibration tests for two sets of mock-ups simulating actual piping in the plants by sinusoidal oscillation and by that obtained at an actual plant were carried out, and then an evaluation method was developed to obtain proper value of vibration stress from the measured data by the vibration tests. In comparison of the vibration stress obtained from the measured acceleration with that directly measured using strain gauges, it is confirmed that accurate vibration stress can be evaluated by a formula in which the real center of gravity of small bore piping and the acceleration of main (system) piping are considered. (author)

Chemical cleaning the service water system at a nuclear power plant

International Nuclear Information System (INIS)

Brice, T.O.; Glover, W.A.

1994-01-01

Chemical cleaning a large cooling water system in a nuclear power plant presented many unique problems. The selection, qualification, and performance of the cleaning process were done using a phased approach. The piping was inspected to determine the extent of the problem. Deposit samples were removed from the service water system pipe and tested in the laboratory to determine the most effective cleaning solvent for deposit removal. An engineering study was performed to define the design parameters required to implement the system-wide chemical cleaning

Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

Energy Technology Data Exchange (ETDEWEB)

Nyman, R; Erixon, S; Tomic, B; Lydell, B

1995-12-01

SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs.

Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

International Nuclear Information System (INIS)

Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

1995-12-01

SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs

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

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)

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.)

Plastic collapse moment for pipe repaired with weld overlay

International Nuclear Information System (INIS)

Li, Yinsheng; Hasegawa, Kunio; Shibuya, Akira; Deardorff, Arthur

2009-01-01

The Weld Overlay has been used in several countries as an effective method to repair the stress corrosion cracks in nuclear power plant piping. However, the method to evaluate the plastic collapse stress for the pipe repaired with Weld Overlay has not been proposed and the limit load criterion for single uniform material has been used to design its structure by now. In this paper, the equations to evaluate the plastic collapse moment for the pipe repaired with Weld Overlay have been derived considering two layer materials. Moreover, several numerical examples are given to show the validity of Weld Overlay. The equations given in this paper are simple to use like the limit load criterion showed in present standards such as JSME Rules on Fitness-for-Service for Nuclear Power Plants or ASME Boiler and Pressure Vessel Code Section XI, and they can not only be used to evaluate the fracture of the pipe, but also be applied to design the weld structure. (author)

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

Mechanical assessment of local thinned pipings

International Nuclear Information System (INIS)

Meister, E.

2007-01-01

Local wall thinning is likely to be found in some piping systems of nuclear power plant under, for example, Flow Accelerated Corrosion in raw water systems or by loss of metal during the grinding of the weld seam. To assess the mechanical integrity in such situations, EDF/SEPTEN has developed calculation methods for the RSE-M (In Service Inspection Rules for the Mechanical components of PWR nuclear power islands) code. This paper focuses on the methodology used for internal pressure resistance evaluation based on limit load calculations. Beyond the Nuclear Safety classification and requirements given by the RSE-M code, this problem is general for Power Piping and the associated in service rules. (author) [fr

The evolution of the break preclusion concept for nuclear power plants in Germany

Energy Technology Data Exchange (ETDEWEB)

Schulz, H. [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

1997-04-01

In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A = open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.

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)

The Development of a Snubber Management System for Welds in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Yoo, Hyun Ju; Cho, Yong-Bae; Kim, Yoo Sung [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

There are snubbers, spring hanger, anchors, rigid supports for the structures which support the static and/or dynamic loads such as thermal load, pressure, impact and vibration from components and pipings of nuclear power plants. Snubbers constrain the displacements generated by loads transmitted to components and systems abruptly. When the loss of function of snubber during normal operation, the thermal load and pressure transmit directly to structures such as pipings and components, and additional loads, which were not considered during the design stage, act on the structures. Therefore, according to regulatory requirement to confirm the stability of supporting system, the inspection for the snubber is reflected to in-service inspection (ISI) and in-service test (IST) plans. In order to comply with the regulatory requirement, KHNP has performed the ISI and IST and inspected the snubbers. As the increment of operating year of nuclear power plants in Korea, the possibility of deterioration of equipment is higher. Therefore, the security related to the integrity of equipment becomes more important. The snubber takes an important role related to the structural integrity equipped on principal pipings. 100% snubbers are inspected during pre-service inspection and 10% snubbers are inspected during in-service inspection as a sample in nuclear power plants in Korea. KHNP has been developed a snubber management system because there was no management tool for snubbers to show the inspection and maintenance results systematically. The inspection and maintenance results of snubbers can be easily reached by plants, head office and CRI. Moreover, the information related to inspection history and condition of snubber can be effectively inquired.

BNL NONLINEAR PRE TEST SEISMIC ANALYSIS FOR THE NUPEC ULTIMATE STRENGTH PIPING TEST PROGRAM

International Nuclear Information System (INIS)

DEGRASSI, G.; HOFMAYER, C.; MURPHY, C.; SUZUKI, K.; NAMITA, Y.

2003-01-01

The Nuclear Power Engineering Corporation (NUPEC) of Japan has been conducting a multi-year research program to investigate the behavior of nuclear power plant piping systems under large seismic loads. The objectives of the program are: to develop a better understanding of the elasto-plastic response and ultimate strength of nuclear piping; to ascertain the seismic safety margin of current piping design codes; and to assess new piping code allowable stress rules. Under this program, NUPEC has performed a large-scale seismic proving test of a representative nuclear power plant piping system. In support of the proving test, a series of materials tests, static and dynamic piping component tests, and seismic tests of simplified piping systems have also been performed. As part of collaborative efforts between the United States and Japan on seismic issues, the US Nuclear Regulatory Commission (USNRC) and its contractor, the Brookhaven National Laboratory (BNL), are participating in this research program by performing pre-test and post-test analyses, and by evaluating the significance of the program results with regard to safety margins. This paper describes BNL's pre-test analysis to predict the elasto-plastic response for one of NUPEC's simplified piping system seismic tests. The capability to simulate the anticipated ratcheting response of the system was of particular interest. Analyses were performed using classical bilinear and multilinear kinematic hardening models as well as a nonlinear kinematic hardening model. Comparisons of analysis results for each plasticity model against test results for a static cycling elbow component test and for a simplified piping system seismic test are presented in the paper

Analysis of failed nuclear plant components

International Nuclear Information System (INIS)

Diercks, D.R.

1993-01-01

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

Analysis of failed nuclear plant components

International Nuclear Information System (INIS)

Diercks, D.R.

1992-07-01

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

Analysis of failure events for expansion joints in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Sato, Masahiro [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Although a large number of expansion joints are used in nuclear power plants with light water reactors, their failure events have not been paid as much attention as those of vessels and pipes. However, as the operation period of nuclear power plants becomes longer, it is necessary to pay attention to their failure events as well as those of vessels and pipes, because aging problems and latent troubles originated in design or fabrication of expansion joints may appear during their long period operation. In this work, we investigated failure event reports of expansion joints in nuclear power plants both in Japan and in U.S.A. and analyzed (1) the influence to output power level, (2) the position and (3) the cause of each failure. It is revealed that the failure events of expansion joints have continuously occurred, some of which have exerted influence upon power level and have caused fatal or injury accidents of personnel, and hence the importance of corrective actions to prevent the recurrence of such events is pointed out. The importance of countermeasures to the following individual events is also pointed out: (1) corrosion of expansion joints in service water systems, (2) degradation of rubber expansion joints in main condensers, (3) vibration and fatigue of expansion joints in extraction steam lines and (4) transgranular stress corrosion cracking of penetration bellows of containments. (author)

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)

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)

Online condition monitoring to enable extended operation of nuclear power plants

International Nuclear Information System (INIS)

Meyer, Ryan Michael; Bond, Leonard John; Ramuhalli, Pradeep

2012-01-01

Safe, secure, and economic operation of nuclear power plants will remain of strategic significance. New and improved monitoring will likely have increased significance in the post-Fukushima world. Prior to Fukushima, many activities were already underway globally to facilitate operation of nuclear power plants beyond their initial licensing periods. Decisions to shut down a nuclear power plant are mostly driven by economic considerations. Online condition monitoring is a means to improve both the safety and economics of extending the operating lifetimes of nuclear power plants, enabling adoption of proactive aging management. With regard to active components (e.g., pumps, valves, motors, etc.), significant experience in other industries has been leveraged to build the science base to support adoption of online condition-based maintenance and proactive aging management in the nuclear industry. Many of the research needs are associated with enabling proactive management of aging in passive components (e.g., pipes, vessels, cables, containment structures, etc.). This paper provides an overview of online condition monitoring for the nuclear power industry with an emphasis on passive components. Following the overview, several technology/knowledge gaps are identified, which require addressing to facilitate widespread online condition monitoring of passive components. (author)

Actions concerning nuclear power plant life evaluation

International Nuclear Information System (INIS)

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

1998-01-01

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

Investigations into the cause of the oil fire in the nuclear power plant, Muehleberg

International Nuclear Information System (INIS)

Hagn, L.; Huppmann, H.

1972-01-01

At the end of July, 1971, an oil fire broke out in the turbine hall of the Swiss Nuclear Power Plant Muehleberg. The cause of the escape of oil from a pressure-oil pipe was the loosening of a cutting ring screwed pipe connection on the servo-motor of a control valve. Material and vibration tests on the main components showed the evolution of the damage to be the combined influences of two factors, viz. 1. vibration stress on the piping and screw joints - dependent on the operational state of the turbine - and resultant possible loosening of the screw caps; 2. loss of the vibration absorbant clip fastenings of the piping. Undetected deformations and dislocations in the screwed pipe connections ensuing from erection were presumably contributory influences. The oil escaping from the leak ignited by autoxidation in the sprayed-on asbestos insulation of the valve housing. From the results of these investigations and the analyses of other oil fires in steam turbine plants insured with the Allianz, recommendations ensue for the prevention and minimising of oil fire damage. (orig.) [de

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)

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)

Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)

2015-05-15

Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by

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

Energy Technology Data Exchange (ETDEWEB)

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

1983-03-01

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

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

International Nuclear Information System (INIS)

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

1983-03-01

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

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.

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.

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

Development of seismic design method for piping system supported by elastoplastic damper. 3. Vibration test of three-dimensional piping model and its response analysis

International Nuclear Information System (INIS)

Namita, Yoshio; Kawahata, Jun-ichi; Ichihashi, Ichiro; Fukuda, Toshihiko.

1995-01-01

Component and piping systems in current nuclear power plants and chemical plants are designed to employ many supports to maintain safety and reliability against earthquakes. However, these supports are rigid and have a slight energy-dissipating effect. It is well known that applying high-damping supports to the piping system is very effective for reducing the seismic response. In this study, we investigated the design method of the elastoplastic damper [energy absorber (EAB)] and the seismic design method for a piping system supported by the EAB. Our final goal is to develop technology for applying the EAB to the piping system of an actual plant. In this paper, the vibration test results of the three-dimensional piping model are presented. From the test results, it is confirmed that EAB has a large energy-dissipating effect and is effective in reducing the seismic response of the piping system, and that the seismic design method for the piping system, which is the response spectrum mode superposition method using each modal damping and requires iterative calculation of EAB displacement, is applicable for the three-dimensional piping model. (author)

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.)

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.

Development of engineering program for integrity evaluation of pipes with local wall thinned defects

International Nuclear Information System (INIS)

Park, Chi Yong; Lee, Sung Ho; Kim, Tae Ryong; Park, Sang Kyu

2008-01-01

Integrity evaluation of pipes with local wall thinning by erosion and corrosion is increasingly important in maintenance of wall thinned carbon steel pipes in nuclear power plants. Though a few program for integrity assessment of wall thinned pipes have been developed in domestic nuclear field, however those are limited to straight pipes and methodology proposed in ASME Sec.XI Code Case N-597. Recently, the engineering program for integrity evaluation of pipes with all kinds of local wall defects such as straight, elbow, reducer and branch pipes was developed successfully. The program was designated as PiTEP (Pipe Thinning Evaluation Program), which name was registered as a trademark in the Korea Intellectual Property Office. A developed program is carried out by sequential step of four integrity evaluation methodologies, which are composed of construction code, code case N-597, its engineering method and two developed owner evaluation method. As PiTEP program will be performed through GUI (Graphic User Interface) with user's familiarity, it would be conveniently used by plant engineers with only measured thickness data, basic operation conditions and pipe data

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)

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

International Nuclear Information System (INIS)

Bronzen, S.

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

Development of computer-aided design and production system for nuclear power plant

International Nuclear Information System (INIS)

Ishii, Masanori

1983-01-01

The technically required matters related to the design and production of nuclear power stations tended to increase from the viewpoint of the safety and reliability, and it is indispensable to cope with such technically required matters skillfully for the rationalization of the design and production and for the construction of highly reliable plants. Ishikawajima Harima Heavy Industries Co., Ltd., has developed the computer-aided design data information and engineering system which performs dialogue type design and drawing, and as the result, the design-production consistent system is developed to do stress analysis, production design, production management and the output of data for numerically controlled machine tools consistently. In this paper, mainly the consistent system in the field of plant design centering around piping and also the computer system for the design of vessels and others are outlined. The features of the design works for nuclear power plants, the rationalization of the design and production management of piping and vessels, and the application of the CAD system to other general equipment and improvement works are reported. This system is the powerful means to meet the requirement of heightening quality and reducing cost. (Kako, I.)

Pipe failure probability - the Thomas paper revisited

International Nuclear Information System (INIS)

Lydell, B.O.Y.

2000-01-01

Almost twenty years ago, in Volume 2 of Reliability Engineering (the predecessor of Reliability Engineering and System Safety), a paper by H. M. Thomas of Rolls Royce and Associates Ltd. presented a generalized approach to the estimation of piping and vessel failure probability. The 'Thomas-approach' used insights from actual failure statistics to calculate the probability of leakage and conditional probability of rupture given leakage. It was intended for practitioners without access to data on the service experience with piping and piping system components. This article revisits the Thomas paper by drawing on insights from development of a new database on piping failures in commercial nuclear power plants worldwide (SKI-PIPE). Partially sponsored by the Swedish Nuclear Power Inspectorate (SKI), the R and D leading up to this note was performed during 1994-1999. Motivated by data requirements of reliability analysis and probabilistic safety assessment (PSA), the new database supports statistical analysis of piping failure data. Against the background of this database development program, the article reviews the applicability of the 'Thomas approach' in applied risk and reliability analysis. It addresses the question whether a new and expanded database on the service experience with piping systems would alter the original piping reliability correlation as suggested by H. M. Thomas

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).

PWR composite materials use. A particular case of safety-related service water pipes

International Nuclear Information System (INIS)

Pays, M.F.; Le Courtois, T.

1997-11-01

This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during 'lifetime'); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author)

PWR composite materials use. A particular case of safety-related service water pipes

Energy Technology Data Exchange (ETDEWEB)

Pays, M.F.; Le Courtois, T

1997-11-01

This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

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

Development of piping evaluation diagram for LBB application to KNGR surge line

International Nuclear Information System (INIS)

Yoon, K. S.; Park, W. B.; Kim, J. M.; Choi, T. S.; Yang, J. S.; Park, C. Y.

1998-01-01

Plant specific data, such as pipe geometry, material properties and pipe loads, are required in order to evaluate Leak-Before-Break (LBB) applicability to piping systems in nuclear power plant under the construction. However, the existing method of LBB evaluation for KSNP's can not be used for newly developed nuclear plants such as Korean Next Generation Reactor (KNGR) which material properties is not available and LBB evaluation is required during design process. In order to solve this problem during developing process for KNGR surge line LBB Piping Evaluation Diagram (PED), which is independent of piping geometry and has a function of the loads applied in piping system, is developed in this paper. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the PED. The PED, therefore, can be used for quick LBB evaluation of KNGR surge line in the process of both design and construction. The benefit obtained by using the PED is : 1) to be able to very quickly confirm LBB applicability without calculating any leakage crack length for all concerned piping locations in the process of both iterative design for optimal routing and construction and 2) to save significantly a lot of computing times required for the corresponding LBB analyses

Assessment of Pipe Wall Loss Using Guided Wave Testing

International Nuclear Information System (INIS)

Joo, Kyung Mun; Jin, Seuk Hong; Moon, Yong Sig

2010-01-01

Flow accelerated corrosion(FAC) of carbon steel pipes in nuclear power plants has been known as one of the major degradation mechanisms. It could have bad influence on the plant reliability and safety. Also detection of FAC is a significant cost to the nuclear power plant because of the need to remove and replace insulation. Recently, the interest of the guided wave testing(GWT) has grown because it allows long range inspection without removing insulation of the pipe except at the probe position. If GWT can be applied to detection of FAC damages, it will can significantly reduce the cost for the inspection of the pipes. The objective of this study was to determine the capability of GWT to identify location of FAC damages. In this paper, three kinds of techniques were used to measure the amplitude ratio between the first and the second welds at the elbow area of mock-ups that contain real FAC damages. As a result, optimal inspection technique and minimum detectability to detect FAC damages drew a conclusion

Specialist meeting on leak before break in reactor piping and vessels

Energy Technology Data Exchange (ETDEWEB)

Bartholome, G.; Bazant, E.; Wellein, R. [Siemens KWU, Stuttgart (Germany)] [and others

1997-04-01

A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

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.

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

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

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.

Leak before break piping evaluation diagram

International Nuclear Information System (INIS)

Fabi, R.J.; Peck, D.A.

1994-01-01

Traditionally Leak Before Break (LBB) has been applied to the evaluation of piping in existing nuclear plants. This paper presents a simple method for evaluating piping systems for LBB during the design process. This method produces a piping evaluation diagram (PED) which defines the LBB requirements to the piping designer for use during the design process. Several sets of LBB analyses are performed for each different pipe size and material considered in the LBB application. The results of this method are independent of the actual pipe routing. Two complete LBB evaluations are performed to determine the maximum allowable stability load, one evaluation for a low normal operating load, and the other evaluation for a high normal operating load. These normal operating loads span the typical loads for the particular system being evaluated. In developing the allowable loads, the appropriate LBB margins are included in the PED preparation. The resulting LBB solutions are plotted as a set of allowable curves for the maximum design basis load, such is the seismic load versus the normal operating load. Since the required margins are already accounted for in the LBB PED, the piping designer can use the diagram directly with the results of the piping analysis and determine immediately if the current piping arrangement passes LBB. Since the LBB PED is independent of pipe routing, changes to the piping system can be evaluated using the existing PED. For a particular application, all that remains is to confirm that the actual materials and pipe sizes assumed in creating the particular design are built into the plant

Water management of the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Rabusic, P.

1990-01-01

Industrial water for the Dukovany nuclear power plant is taken from a reservoir built on the Jihlava river. The volume of the reservoir is 17.1 million m 3 . The water taken serves mainly as the make-up water for the cooling circuit and is treated by clearing; it is also used for the preparation of demineralized water for making up the condensate circuit and for the preparation of soft water for making up the heat piping. The consumption of industrial water is 1 to 2 m 3 /s. Waste waters are released into rainwater drainage, sewage drainage and industrial water drainage. Waste waters are segregated according to the place of their origin and are purified, mainly on ion exchangers and on an evaporator. In normal conditions, they are returned to the operation. Concentrated liquid residues and solid wastes will be stored by using the multiple containment system. The most important radioisotopes that may be present in nuclear power plant waste waters, the water and radionuclide balance of the Dukovany nuclear power plant, and chemical and radiological data on the Dukovany waste waters (1989) are tabulated. (P.A.). 6 refs

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

Vibration test on KMRR reactor structure and primary cooling system piping

Energy Technology Data Exchange (ETDEWEB)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

Vibration test on KMRR reactor structure and primary cooling system piping

International Nuclear Information System (INIS)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

Nuclear regulation. License renewal questions for nuclear plants need to be resolved

International Nuclear Information System (INIS)

Fultz, Keith O.; Kruslicky, Mary Ann; McDowell, William D. Jr.; Coleman, Robert L.

1989-04-01

A December 1986 pipe rupture at Virginia Power's Surry unit 2 nuclear power plant injured eight workers; four later died. As a result of this accident, Representative Edward J. Markey requested GAO to examine the Surry accident and assess the problems confronting aging nuclear plants. In March 1988 we reported our findings concerning the accident and a July 1987 incident at the Trojan nuclear plant in Oregon. This report addresses problems confronting aging nuclear plants by examining the Nuclear Regulatory Commission's (NRC) program to develop a license renewal policy and accompanying regulations, and the initiatives underway by the Department of Energy (DOE) and the electric utility industry to extend the operating lives of these plants. Nuclear power has become second only to coal as the largest producer of electricity in the United States. The 110 nuclear plants currently in service are operated by 54 utilities, provide about 20 percent of the nation's electricity, and represent a capital investment of over $200 billion. The Atomic Energy Act authorizes NRC to issue nuclear plant operating licenses for up to 40 years and provides for license extensions beyond the initial operating period. The act does not, however, stipulate the criteria for evaluating a utility request to operate a nuclear plant longer than 40 years. The oldest operating license currently in effect will expire in the year 2000. According to NRC, about one-half of the existing operating licenses will terminate by the year 2015, and most licenses will expire by about 2030. Many utilities will have to decide in the early 1990s whether to continue operating older nuclear plants or to construct new generating capacity. A clear understanding of the terms and conditions governing the license renewal process will be a key element in deciding how to meet future electricity demand. Although NRC has developed 3 possible license renewal policy options and identified 15 areas of regulatory uncertainty that

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

Functional capability of piping systems

International Nuclear Information System (INIS)

Terao, D.; Rodabaugh, E.C.

1992-11-01

General Design Criterion I of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations requires, in part, that structures, systems, and components important to safety be designed to withstand the effects of earthquakes without a loss of capability to perform their safety function. ne function of a piping system is to convey fluids from one location to another. The functional capability of a piping system might be lost if, for example, the cross-sectional flow area of the pipe were deformed to such an extent that the required flow through the pipe would be restricted. The objective of this report is to examine the present rules in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, and potential changes to these rules, to determine if they are adequate for ensuring the functional capability of safety-related piping systems in nuclear power plants

Promethus Hot Leg Piping Concept

International Nuclear Information System (INIS)

AM Girbik; PA Dilorenzo

2006-01-01

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept

Monitoring corrosion and biofilm formation in nuclear plants using electrochemical methods

International Nuclear Information System (INIS)

Licina, G.J.; Nekoksa, G.; Ward, G.L.; Howard, R.L.; Cubicciotti, D.

1993-01-01

During the 1980's, degradation of piping, heat exchangers, and other components in raw water cooled systems by a variety of corrosion mechanisms became an important in the reliability and cost effectiveness of U.S. nuclear plants. General and localized corrosion, including pitting and crevice corrosion, have all been shown to be operative in nuclear plant cooling systems. Microbiologically influenced corrosion (MIC) also afflicts nuclear cooling water and service water systems. The prediction of locations to be inspected, selection of mitigation measures, and control of water treatments and maintenance planning rely upon the accuracy and sensitivity of monitoring techniques. Electrochemical methods can provide rapid measurements of corrosion and biological activity on line. The results from a corrosion monitoring study in a service water system at a fresh water cooled nuclear plant are presented. This study utilized determinations of open circuit potential and reversed potentiodynamic scans on carbon steels, Admiralty, and stainless steels (Types 304 and 316 as well as high chromium, high molybdenum ferritic and austenitic grades) to evaluate the rate and form of corrosion to be anticipated in typical service. An electrochemical method that permits the monitoring of biofilm activity on-line has been developed. Results from laboratory and in-plant exposure in a nuclear power plant system are presented

Preliminary Study for Development of Welds Integrity Verification Equipment for the Small Bore Piping

International Nuclear Information System (INIS)

Choi, Geun Suk; Lee, Jong Eun; Ryu, Jung Hoon; Cho, Kyoung Youn; Sohn, Myoung Sung; Lee, Sanghoon; Sung, Gi Ho; Cho, Hong Seok

2016-01-01

It has been reported leakage accident of small-bore piping in Korea. Leakage accident of small-bore pipes are those that will increase due to the aging of the nuclear power plant. And if leakage of the pipe is repaired by using the clamping device when it occur accident, it is economically benefits. The clamping device is a fastening device used to hold or secure objects tightly together to prevent movement or separation through the application of inward pressure. However, when the accident occurs, it can't immediately respond because maintenance and repairing technology are not institutionalized in KEPIC. Thus it appears an economic loss. The technology for corresponding thereto is necessary for the safety of the operation of nuclear power plants. The purpose of this research is to develop an online repairing technology of socket welded pipe and vibration monitoring system of small-bore pipe in the nuclear power plant. Specifically, detailed studies are as follows : • Development of weld overlay method of safety class socket welded connections • Development of Mechanical Clamping Devices for Safety Class 2, 3 small-bore pipe. The purpose of this study is to develop an online repairing technology of socket welded pipe and vibration monitoring system of small-bore pipe, resulting in degraded plant systems. And it is necessary to institutionalize the technology. The fatigue crack testing of socket welded overlay will be performed and fatigue life evaluation method will be developed in second year. Also prototype fabrication of mechanical clamping device will be completed. Base on final goal, the intent is to propose practical evaluation tools, design and fabrication methods for socket welded connection integrity. And result of this study is to development of KEPIC code case approved technology for on-line repairing system of socket welded connection and fabrication of mechanical clamping device

Survey of extreme load design regulatory agency licensing requirements for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Stevenson, J D

1976-04-01

Since 1965, when extreme load requirements began to be considered explicitly in nuclear power plant design, there has been a gradual divergence in requirements imposed by national regulatory agencies. However, nuclear plant safety is an international problem because of the potential international effects of any postulated plant failure. For this reason this paper has been prepared in an attempt to highlight the differences in national criteria currently used in the extreme load design of nuclear plant facilities. No attempt has been made to evaluate the relative merit of the criteria established by the various national regulatory agencies. This paper presents the results of a recent survey made of national atomic energy regulatory agencies and major nuclear steam supply design agencies, which requested a summary of current licensing criteria associated with earthquake, extreme wind (tornado), flood, airplane crash and accident (pipe break) loads applicable within the various national jurisdictions. Also presented are a number of comparisons which are meant to illustrate the differences in national regulatory criteria.

Survey of extreme load design regulatory agency licensing requirements for nuclear power plants

International Nuclear Information System (INIS)

Stevenson, J.D.

1976-01-01

Since 1965, when extreme load requirements began to be considered explicitly in nuclear power plant design, there has been a gradual divergence in requirements imposed by national regulatory agencies. However, nuclear plant safety is an international problem because of the potential international effects of any postulated plant failure. For this reason this paper has been prepared in an attempt to highlight the differences in national criteria currently used in the extreme load design of nuclear plant facilities. No attempt has been made to evaluate the relative merit of the criteria established by the various national regulatory agencies. This paper presents the results of a recent survey made of national atomic energy regulatory agencies and major nuclear steam supply design agencies, which requested a summary of current licensing criteria associated with earthquake, extreme wind (tornado), flood, airplane crash and accident (pipe break) loads applicable within the various national jurisdictions. Also presented are a number of comparisons which are meant to illustrate the differences in national regulatory criteria. (Auth.)

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

Application of the SSMRP methodology to the seismic risk at the Zion Nuclear Power Plant

International Nuclear Information System (INIS)

Bohn, M.P.; Shieh, L.C.; Wells, J.E.

1983-11-01

The Seismic Safety Margins Research Program (SSMRP) has the goal of developing a fully coupled analysis procedure for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. The risk analysis included a detailed seismological evaluation of the region around Zion, Illinois, which provided the earthquake-hazard function and a randomized set of 180 time histories (having peak ground acceleration values up to 1.8 g). These time histories were used as input for dynamic structural response calculations for four different Zion buildings. Detailed finite-element models of the buildings were used. Calculated time histories at piping support points were then used to determine moments throughout critical piping systems. Twenty separate piping models were analyzed. Finally, the responses of piping and safety system components within the buildings were combined with probabilistic failure criteria and event-tree/fault-tree models of the plant safety systems to produce an estimate of the frequency of core melt and radioactive release due to earthquakes

Shock resistance of composite material pipes

International Nuclear Information System (INIS)

Pays, M.F.

1995-01-01

Composite materials have found a wide range of applications for EDF nuclear plants. Applications include fire pipework, demineralized water, service water, and emergency-supplied service water piping. Some of those pipework is classified nuclear safety, their integrity (resistance to water aging and earthquakes or accidental excess pressure (water hammer)) must be safeguarded. As composite materials generally suffer damage for low energy impacts (under 10 J), the pipes planned for the Civaux power plant have been studied for their resistance to a low speed shock (0 to 50 m/s) and of a 0 to 110 J energy level. For three representative diameters (20, 150, 600 mm), the minimum impact energy that leads to a leak has been determined to be respectively 18, 20 and 48 J. Then the leak rate versus impact energy was plotted; until roughly 90 J, the leak rate remains stable at less than 25 cm 3 /h and raises to higher values (300 cm 3 /h) afterwards. The level of leakage in the range of impact energy tested always stays within the limits set by the Safety Authorities for metallic pipes. These results have been linked to destructive examinations, to clarify the damage mechanisms. Other tests are still ongoing to follow the evolution of the damage and of the leak rate while the pipe is maintained under service pressure during one year

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

Hot Leg Piping Materials Issues

International Nuclear Information System (INIS)

V. Munne

2006-01-01

With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the space nuclear power plant (SNPP) for Project Prometheus (References a and b) the reactor outlet piping was recognized to require a design that utilizes internal insulation (Reference c). The initial pipe design suggested ceramic fiber blanket as the insulation material based on requirements associated with service temperature capability within the expected range, very low thermal conductivity, and low density. Nevertheless, it was not considered to be well suited for internal insulation use because its very high surface area and proclivity for holding adsorbed gases, especially water, would make outgassing a source of contaminant gases in the He-Xe working fluid. Additionally, ceramic fiber blanket insulating materials become very friable after relatively short service periods at working temperatures and small pieces of fiber could be dislodged and contaminate the system. Consequently, alternative insulation materials were sought that would have comparable thermal properties and density but superior structural integrity and greatly reduced outgassing. This letter provides technical information regarding insulation and materials issues for the Hot Leg Piping preconceptual design developed for the Project Prometheus space nuclear power plant (SNPP)

Field measurement of the piping system vibration of Ko-Ri unit 4 during the load-following operation

International Nuclear Information System (INIS)

Chung, Tae-Young; Hong, Sung-Yull; Kim, Bum-Nyun.

1989-01-01

During the load-following operation of nuclear power plants, flow rate, temperature, and pressure in the piping system can be varied by changing the electric power output level, and these variations can cause different vibration phenomena in the piping system. The piping system vibration is important because it is directly related to the dynamic stress of the piping system and can affect the life of the piping system through structural fatigue. An assessment of vibration levels for the classes II and III piping systems of the Ko-Ri Unit 4950-MW nuclear power plant was performed according to the given pattern of the load-following operation to study its feasibility from the viewpoint of piping system vibration. The classes II and III piping system vibration of the Ko-Ri Unit 4 may not cause any potential problem under the given pattern of the load-following operation; however, it is recommended that long-term operation in the 85 to 95% power range be avoided as much as possible

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.)

International piping integrity research group (IPIRG) program final report

International Nuclear Information System (INIS)

Schmidt, R.; Wilkowski, G.; Scott, P.; Olsen, R.; Marschall, C.; Vieth, P.; Paul, D.

1992-04-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Programme. The IPIRG Programme was an international group programme managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United states. The objective of the programme was to develop data needed to verify engineering methods for assessing the integrity of nuclear power plant piping that contains circumferential defects. The primary focus was an experimental task that investigated the behaviour of circumferentially flawed piping and piping systems to high-rate loading typical of seismic events. To accomplish these objectives a unique pipe loop test facility was designed and constructed. The pipe system was an expansion loop with over 30 m of 406-mm diameter pipe and five long radius elbows. Five experiments on flawed piping were conducted to failure in this facility with dynamic excitation. The report: provides background information on leak-before-break and flaw evaluation procedures in piping; summarizes the technical results of the programme; gives a relatively detailed assessment of the results from the various pipe fracture experiments and complementary analyses; and, summarizes the advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG Program

Pipe fracture evaluations for leak-rate detection: Probabilistic models

International Nuclear Information System (INIS)

Rahman, S.; Wilkowski, G.; Ghadiali, N.

1993-01-01

This is the second in series of three papers generated from studies on nuclear pipe fracture evaluations for leak-rate detection. This paper focuses on the development of novel probabilistic models for stochastic performance evaluation of degraded nuclear piping systems. It was accomplished here in three distinct stages. First, a statistical analysis was conducted to characterize various input variables for thermo-hydraulic analysis and elastic-plastic fracture mechanics, such as material properties of pipe, crack morphology variables, and location of cracks found in nuclear piping. Second, a new stochastic model was developed to evaluate performance of degraded piping systems. It is based on accurate deterministic models for thermo-hydraulic and fracture mechanics analyses described in the first paper, statistical characterization of various input variables, and state-of-the-art methods of modem structural reliability theory. From this model. the conditional probability of failure as a function of leak-rate detection capability of the piping systems can be predicted. Third, a numerical example was presented to illustrate the proposed model for piping reliability analyses. Results clearly showed that the model provides satisfactory estimates of conditional failure probability with much less computational effort when compared with those obtained from Monte Carlo simulation. The probabilistic model developed in this paper will be applied to various piping in boiling water reactor and pressurized water reactor plants for leak-rate detection applications

Proceedings of the topical meeting on nuclear power plant life extension

International Nuclear Information System (INIS)

Anon.

1988-01-01

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

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.)

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

Establishment of the Integrated Data of Mechanical and Physical Properties of Nuclear Structural Materials for the Safety of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Ho; Maeng, Wan Young; Kim, Woogun; and others

2012-01-15

It is essential to have a evaluation mock system for integrity verification of operating nuclear power plants. There is no material evaluation facility in Korea with operating conditions. It is needed to develop a large scale materials and components integrity evaluation mock system for a design, construction, and operation of Korean nuclear power plants. Improvement of nuclear technology in materials design, construction, and operation Korea had a chance to export nuclear power plant to other country. In order to keep a competitive power in the nuclear industries, it is important to set up a materials and components integrity evaluation system. Standard or code of materials and component design and operation is also essential for keeping the initiative in the nuclear industries. It takes a long period of time under simulated conditions close to the conditions of nuclear power plants (NPPs), to have flow accelerated corrosion (FAC) tests for the piping in their secondary system. Therefore, it will be efficient to design the multipurpose test section. A device to monitor electrochemical properties continuously is installed to evaluate environmental conditions for FAC of the piping. The flow velocity of the loop is controlled by a valve under high temperature and high pressure conditions, since the measured electrochemical value is influenced by the flow velocity. Dummy sensors are installed to accelerate the FAC by making turbulence of the flow. It is designed to test the piping with different diameters and of alternative modified materials, at the same time. The specimens are checked during or after each test by destruct ve and non-destructive methods. And small rotating specimens will be tested in an autoclave to measure the very small amount of weight change, when the thickness change is not significant. A test loop has been designed to check electrochemical properties and it can control flow velocity, dissolved oxygen, pH and temperature of the coolant. The

Thinned pipe management program of Korean NPPs

International Nuclear Information System (INIS)

Lee, S.H.; Kim, T.R.; Jeon, S.C.; Hwang, K.M.

2003-01-01

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle systems in Nuclear Power Plants (NPP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain stress and consequently results in leakage or rupture. In order to minimize the possibility of excessive wall thinning, Thinned Pipe Management Program (TPMP) has been set up and being implemented to all Korean NPPs. Important elements of the TPMP include the prediction of the FAC rate for each component based on model analysis, prioritization of pipe components for inspection, thickness measurement, calculation of wear and wear rate for each component. Additionally, decision making associated with replacement or continuous service for thinned pipe components and establishment of long-term strategic management plan based on diagnosis of plant condition regarding overall wall thinning also are essential part of the TPMP. From pre-service inspection data, it has been found that initial thickness is varies, which influences wear and wear rate calculations. (author)

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

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)

Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

International Nuclear Information System (INIS)

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

2015-01-01

Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

Remotely controlled repair of piping at Douglas Point

International Nuclear Information System (INIS)

Conrath, J.J.

1983-06-01

The 200 MWe Douglas Point Nuclear Generating Station which started operation in 1966 was Canada's first commercial nuclear power plant. In 1977, after 11 years of operation, leakage of heavy water was detected and traced to the Moderator Piping System (pipe sizes 19 mm to 76 mm) located in a vault below the reactor where the radiation fields during shutdown ranged up to 5000 R/Hr. Inspection using remotely operated TV cameras showed that a 'U' bolt clamp support had worn through the wall of one pipe and resulted in the leakage and also that wear was occurring on other pipes. An extensive repair plan was subsequently undertaken in the form of a joint venture of the designer-owner Atomic Energy of Canada Limited, and the builder-operator, Ontario Hydro. This paper describes the equipment and procedures used in remotely controlled repairs at Douglas Point

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

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

Pipe whip analysis using the TEDEL code

International Nuclear Information System (INIS)

Millard, D.; Hoffmann, A.

1985-02-01

In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented

A compilation of experiences of corrosion in Nordic nuclear power plants

International Nuclear Information System (INIS)

Norring, K.; Rosborg, B.

1985-01-01

14 reactors in commercial operation in the Nordic countries exhibit a great variety of corrosion induced damages. The largest number of such damages have affected turbine plants and seawater cooling systems. More severe cases of corrosion which have been experienced are intergranular stress corrosion cracking of steam generator tubing, stainless steel piping, and high strenght bolts and screws, together with erosion corrosion of structural steel in turbine plants. In all units in operation some form of corrosion damage has occurred. In a worldwide perspective the corrosion problems in the Nordic nuclear power plants have been of manageable extent

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

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...

Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

International Nuclear Information System (INIS)

Wang, G.Y.; Shin, Y.W.; Moody, F.J.

1991-01-01

This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

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

Stress analysis of piping systems and piping supports. Documentation

International Nuclear Information System (INIS)

Rusitschka, Erwin

1999-01-01

The presentation is focused on the Computer Aided Tools and Methods used by Siemens/KWU in the engineering activities for Nuclear Power Plant Design and Service. In the multi-disciplinary environment, KWU has developed specific tools to support As-Built Documentation as well as Service Activities. A special application based on Close Range Photogrammetry (PHOCAS) has been developed to support revamp planning even in a high level radiation environment. It comprises three completely inter-compatible expansion modules - Photo Catalog, Photo Database and 3D-Model - to generate objects which offer progressively more utilization and analysis options. To support the outage planning of NPP/CAD-based tools have been developed. The presentation gives also an overview of the broad range of skills and references in: Plant Layout and Design using 3D-CAD-Tools; evaluation of Earthquake Safety (Seismic Screening); Revamps in Existing Plants; Inter-disciplinary coordination of project engineering and execution fields; Consulting and Assistance; Conceptual Studies; Stress Analysis of Piping Systems and Piping Supports; Documentation; Training and Supports in CAD-Design, etc. All activities are performed to the greatest extent possible using proven data-processing tools. (author)

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)

Development of a simplified piping support system

International Nuclear Information System (INIS)

Leung, J.; Anderson, P.H.; Tang, Y.K.; Kassawara, R.P.; Tang, H.T.

1987-01-01

This paper presents the results of experimental and analytical studies for developing a simplified piping support system (SPSS) for nuclear power piping in place of snubbers. The basic concept of the SPSS is a passive seismic support system consisting of limit stops. Large gaps are provided to allow for free thermal expansion during normal plant operation while preventing excessive displacement during a seismic event. The results are part of a research and development program sponsored by EPRI. (orig./HP)

Experience with diagnostic instrumentation in nuclear power plants

International Nuclear Information System (INIS)

Gopal, R.; Ciaramitaro, W.

1977-01-01

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

Nuclear power plant safety. The merits of separation

International Nuclear Information System (INIS)

Helander, L.I.; Tiren, L.I.

1977-01-01

The paper illustrates how the physical separation of safety-related structures, systems and components can improve the protection of a nuclear power plant against multiple failures that may be caused by events such as fire, pipe-whip, missiles, flooding, hurricanes, lightning etc. Criteria for redundancy and separation requirements affect nuclear plant design in many areas such as building layout, arrangements for fire protection and ventilation, separation of mechanical systems and components, in particular emergency cooling systems, separation of electric equipment and cables. The implementation of the ensuing design criteria for a BWR power plant is described. This design involves the separation of emergency cooling systems into four 50% capacity systems, which are independent and separated, including the distribution network for electric power from on-site standby diesel generators and the circuitry for the reactor protection system. The plant is subdivided into a number of fire zones, each with its own independent ventilation system. The fire zones are further subdivided into a multitude of fire cells such that redundant subsystems are housed in separate cells. These design precautions with regard to fire are complemented by extensive fire fighting systems. (author)

Design and analysis for piping systems

International Nuclear Information System (INIS)

Sterkel, H.-P.; Cutrim, J.H.C.

1981-01-01

The procedure and the typical techniques that are used in NUCLEN for the design and the calculation of the piping of Nuclear Plants. The classification system are generically described and the analysis techniques which are used for the design and verification of the piping systems, i.e. pressure design for the dimensioning of the wallthicknesses, temperature and dead weight analysis together with determination of support points, are shown. The techniques of dynamic design and analyses are described for earthquake and pressure impulse loadings. (Author) [pt

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.)

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)

Mathematical model use for evaluation of radioactivity spreading in nuclear power plant

International Nuclear Information System (INIS)

Kubik, I.; Gladki, Eh.; Yanchik, O.

1976-01-01

On the basis of knowledges of radioactive products behaviour and their spreading in nuclear power plant under normal and accident conditions a KOMPLEX program is developed in the FORTRAN 4 language, permitting to calculate the activity in separate parts of the nuclear power plant with WWR type reactor. The COMPLEX program includes the following subprograms: AZ - PRIM - for estimating active products in fuel, coolant, on the surfaces of fuel element cans and the primary circuit. The subprogram permits to estimate the coolant activity at the expense of fission fragments for 4 different leakage mechanisms: due to diffusion, considerable fuel element damage, contamination of fuel element can surface and fuel washout by coolant; KOR - the program for estimating active corrosion products; ACT - the program for estimating the activity of activation products; CONT - the program for estimating the activity in the nuclear power plant premises (protection envelop) and ventilating pipe. The desciption of the above subprograms is given. For testing of the mathematical model applicability and the possibilities of the corresponding programs the checking calculations for operating parameters of nuclear power plant with WWR type reactor were carried out. The calculation results obtained have shown the applicability of the model suggested and the corresponding programes for nuclear power plant under normal operation and accident conditions [ru

International Piping Integrity Research Group (IPIRG) Program. Final report

International Nuclear Information System (INIS)

Wilkowski, G.; Schmidt, R.; Scott, P.

1997-06-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Program. The IPIRG Program was an international group program managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United States. The program objective was to develop data needed to verify engineering methods for assessing the integrity of circumferentially-cracked nuclear power plant piping. The primary focus was an experimental task that investigated the behavior of circumferentially flawed piping systems subjected to high-rate loadings typical of seismic events. To accomplish these objectives a pipe system fabricated as an expansion loop with over 30 meters of 16-inch diameter pipe and five long radius elbows was constructed. Five dynamic, cyclic, flawed piping experiments were conducted using this facility. This report: (1) provides background information on leak-before-break and flaw evaluation procedures for piping, (2) summarizes technical results of the program, (3) gives a relatively detailed assessment of the results from the pipe fracture experiments and complementary analyses, and (4) summarizes advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG program

International Piping Integrity Research Group (IPIRG) Program. Final report

Energy Technology Data Exchange (ETDEWEB)

Wilkowski, G.; Schmidt, R.; Scott, P. [and others

1997-06-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Program. The IPIRG Program was an international group program managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United States. The program objective was to develop data needed to verify engineering methods for assessing the integrity of circumferentially-cracked nuclear power plant piping. The primary focus was an experimental task that investigated the behavior of circumferentially flawed piping systems subjected to high-rate loadings typical of seismic events. To accomplish these objectives a pipe system fabricated as an expansion loop with over 30 meters of 16-inch diameter pipe and five long radius elbows was constructed. Five dynamic, cyclic, flawed piping experiments were conducted using this facility. This report: (1) provides background information on leak-before-break and flaw evaluation procedures for piping, (2) summarizes technical results of the program, (3) gives a relatively detailed assessment of the results from the pipe fracture experiments and complementary analyses, and (4) summarizes advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG program.

Aging countermeasures for nuclear power plants in Japan and PLEC's activities

International Nuclear Information System (INIS)

Tajima, K.; Ishikawa, M.; Koyama, M.

2002-01-01

on the verification test results of the Structural Assessment of Flawed Equipment (SAF), which was conducted to confirm consistency of behavior between experiments of fracture mechanics and analyses concerning small flaws hypothesized on the nuclear power plant components, in order to verify integrity of the reactor components and piping. SAF model tests were performed to confirm flaw growth and propagation, and also to evaluate safety margin of the actual equipment using small and large test model with small flaws simulating the production pressure vessel and piping. Regarding information service for domestic and overseas, the PLEC plans to carry out these activities by means of international co-operation and sharing of information and experience related to aging. The PLEC collects aging-related information, i.e., incidents, inspection, evaluation methods countermeasures, codes and standards for aging. (author)

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)

Technical Review on Fitness-for-Service for Buried Pipe by ASME Code Case N-806

International Nuclear Information System (INIS)

Park, Sang Kyu; Lee, Yo Seop; So, Il Su; Lim, Bu Taek

2012-01-01

Fitness-for-Service is a useful technology to determine replacement timing, next inspection timing or in-service when nuclear power plant's buried pipes are damaged. If is possible for buried pipes to be aged by material loss, cracks and occlusion as operating time goes by. Therefore Fitness-for-Service technology for buried pipe is useful for plant industry to perform replacement and repair. Fitness-for-Service for buried pipe is studied in terms of existing code and standard for Fitness-for-Service and a current developing code case. Fitness-for-Service for buried pipe was performed according to Code Case N-806 developed by ASME (American Society of Mechanical Engineers)

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)

In-service materials testing of selected components of unit 1 and 2 of V-1 nuclear power plant

International Nuclear Information System (INIS)

Cintula, J.

1982-01-01

The task of in-service nondestructive testing of nuclear installations is to confirm that the state of base material and welded joints has not changed owing to mechanical, thermal or radiation stress. Under the regulations of safe operation the first in-service inspection of all components of a WWER 440 reactor must be carried out after 15,000 to 2O,00O operating hours at the latest. Further in-service inspections are repeated after 30,000 hours (pressure vessels) and 40,000 hours (the main steam piping and the feedwater piping). Proceeding from experience gained so far, intervals are suggested for in-service checks of the other components of the V-1 nuclear power plant. Also briefly described are the main nondestructive methods used for such checks at this power plant. (Z.M.)

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.)

Development of a simplified piping support system

International Nuclear Information System (INIS)

Leung, J.; Anderson, P.H.; Tang, Y.K.; Kassawara, R.P.; Tang, H.T.

1987-01-01

This paper presents the results of experimental and analytical studies for developing a simplified piping support system (SPSS) for nuclear power piping in place of snubbers. The basic concept of the SPSS is a passive seismic support system consisting of limit stops. Large gaps are provided to allow for free thermal expansion during normal plant operation while preventing excessive displacement during a seismic event. The results are part of a research and development program sponsored by the Electric Power Research Institute

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

Developing a program to identify and track corrosion in nuclear plant raw water systems

International Nuclear Information System (INIS)

Spires PE, G.V.; Pickles PE, S.B.

2001-01-01

Findings derived from a comprehensive plant performance survey at Ontario Power Generation's (OPG) nuclear units convinced management that it would be prudent to expand the ongoing power piping Flow Accelerated Corrosion (FAC) induced wall thinning base-lining and tracking program to encompass the raw cooling water systems as well. Such systems are subject to a distinctly different class of pipe wall thinning (PWT) mechanisms than the FAC that degrades high-energy power piping. This paper describes the PWT corrosion assessment and tracking program that has been developed and is currently being implemented by OPG for the raw cooling water (i.e., Service Water) systems within it's nuclear generating stations. Interim databases are used prior to initial inspection rounds to catalogue the prospective locations. For each piping system being surveyed, these interim databases include physical coordinates for the candidate locations, the type and wall thickness of the components comprising each location, ranking indications and recommended NDE methodologies as a function of the anticipated corrosion mechanisms. Rationales for assessing corrosion susceptibility and ranking prospective inspection sites are expounded by way of notations built into the database. (authors)

Experimental investigation of high cycle thermal fatigue in a T-junction piping system

Energy Technology Data Exchange (ETDEWEB)

Selvam, P. Karthick; Kulenovic, Rudi; Laurien, Eckart [Stuttgart Univ. (Germany). Inst. of Nuclear Technology and Energy Systems (IKE)

2015-10-15

High cycle thermal fatigue damage of structure in the vicinity of T-junction piping systems in nuclear power plants is of importance. Mixing of coolant streams at significant temperature differences causes thermal fluctuations near piping wall leading to gradual thermal degradation. Flow mixing in a T-junction is performed. The determined factors result in bending stresses being imposed on the piping system ('Banana effect').

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)

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

Nuclear power plant safety, the merits of separation

International Nuclear Information System (INIS)

Helander, L.I.; Tiren, L.I.

1977-01-01

The United States AEC General Design Criteria for Nuclear Power Plants are used worldwide as a basis for the assessment of nuclear plant safety. Several of these criteria require redundancy of safety systems, separation of protection and control systems, consideration of natural phenomena, etc. All these criteria point in one particular direction: the necessity for physically separating the various safety-related systems of a nuclear power plant, particularly with regard to single occurrences that may yield a multiple failure. Requirements in this regard have been amplified by the United States NRC Regulatory Guides and by IEEE Standards. The single occurrence that yields a multiple failure may be, for example, fire, pipe whip, missiles, flooding, hurricanes, or lightning. The paper discusses protection, against the quoted events and others, obtained through applying criteria regarding redundancy and separation of safety-related structures, systems and components. Such criteria affect nuclear plant design in many areas, such as building lay-out, arrangements for fire protection and ventilation, separation of mechanical systems and components, in particular emergency cooling systems, and separation of electric equipment and cables. Implementation of the ensuing design criteria for a BWR power plant is described. This design involves the separation of Emergency Cooling Systems into four 50% Capacity Systems which are independent and separated, including the distribution network for electric power from on-site standby diesel generators and the circuitry for the reactor protection system. The plant is subdivided into a number of fire zones each with its own independent ventilation system. The fire zones are further subdivided into a multitude of fire cells such that redundant subsystems are housed in separate cells. These design precautions with regard to fire are complemented by extensive fire fighting systems

Control device for the recycling flow rate in a nuclear power plant

International Nuclear Information System (INIS)

Tanigawa, Naoshi; Shida, Toichi.

1983-01-01

Purpose: To prevent the cavitation in a recycling pump even under the conditions where a recycling pump run-back is inhibited, thereby secure the safety of equipments. Constitution: An AND circuit is disposed to a recycling flow rate control system in a BWR type nuclear power plant that issues an output on the condition that a run-back signal is present together with a contact signal of a scoop pipe locking relay or a contact signal of a scoop pipe locking relay. Then, if a demand for the run-back operation of a nuclear reactor recycling pump is issued, the reactor recycling pump is forcedly tripped. Since the pump can always be tripped upon requirement of run-back even if there are some or other inhibitive factors, generation of the cavitation in recycling system equipments can be prevented thereby prevent the damages in the equipments. (Moriyama, K.)

Applied reliability assessment for the passive safety systems of nuclear power plants (NPPs) using system dynamics (SD)

International Nuclear Information System (INIS)

Kim, Yun Il; Woo, Tae Ho

2018-01-01

The passive system by the free-fall is investigated in the accident of nuclear power plants (NPPs). The complex algorithm of the system dynamics (SD) modeling is done in the passive cooling system. The nuclear passive system by free-fall is successfully modeled for the loss of coolant accident (LOCA). Conventional passive system of gravity or natural circulation is working only when the piping systems is in the good condition. The external coolant supply system is introduced in the case of the piping system failure. The water is poured into the reactor through the guiding piping or tube. If the explosion happens, the coolants could be showering into the reactor core and its building. New kind of passive system is expected successfully in the on-site black out where the drone could be operated by battery or engine.

Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

International Nuclear Information System (INIS)

Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

1996-01-01

A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

Microcomputer generated pipe support calculations

International Nuclear Information System (INIS)

Hankinson, R.F.; Czarnowski, P.; Roemer, R.E.

1991-01-01

The cost and complexity of pipe support design has been a continuing challenge to the construction and modification of commercial nuclear facilities. Typically, pipe support design or qualification projects have required large numbers of engineers centrally located with access to mainframe computer facilities. Much engineering time has been spent repetitively performing a sequence of tasks to address complex design criteria and consolidating the results of calculations into documentation packages in accordance with strict quality requirements. The continuing challenges of cost and quality, the need for support engineering services at operating plant sites, and the substantial recent advances in microcomputer systems suggested that a stand-alone microcomputer pipe support calculation generator was feasible and had become a necessity for providing cost-effective and high quality pipe support engineering services to the industry. This paper outlines the preparation for, and the development of, an integrated pipe support design/evaluation software system which maintains all computer programs in the same environment, minimizes manual performance of standard or repetitive tasks, and generates a high quality calculation which is consistent and easily followed

Strain Limits within the Scope of the Integrity Assessment of Piping Systems

International Nuclear Information System (INIS)

Mutz, Alexander

2008-01-01

Allowable stresses in nuclear power plant piping resulting from loading conditions to be considered in Germany are determined on the basis of the German Safety Standards of the Nuclear Safety Standards Commission, KTA. The limitation of the different stress categories within the analysis of the mechanical behaviour is based on a linear elastic material behaviour. Because of the ductile material used in high energy nuclear piping, a more realistic assessment can be performed on the basis of allowable strains using elastic plastic material behaviour. In the present work comparison between the analysis of piping systems considering the elastic material model and the actual elastic plastic material behaviour is performed. The possibilities of allocating plastic strains to calculated elastic stresses is discussed. A parametric study on straight pipes with the actual elastic plastic material model under pure bending is the basis of deriving the elastic plastic strains for the calculated elastic stresses. Strain limits are suggested which correspond to the different stress categories. The aim is to utilize the deformation possibilities of ductile materials used in German nuclear piping and the allocation of maximum strains to different load categories. Keywords: strain limit, ductile material, stress category. (author)

Strain Limits within the Scope of the Integrity Assessment of Piping Systems

Energy Technology Data Exchange (ETDEWEB)

Mutz, Alexander [EnBW, Durlacher Allee 93, Karlsruhe 76131 (Germany)

2008-07-01

Allowable stresses in nuclear power plant piping resulting from loading conditions to be considered in Germany are determined on the basis of the German Safety Standards of the Nuclear Safety Standards Commission, KTA. The limitation of the different stress categories within the analysis of the mechanical behaviour is based on a linear elastic material behaviour. Because of the ductile material used in high energy nuclear piping, a more realistic assessment can be performed on the basis of allowable strains using elastic plastic material behaviour. In the present work comparison between the analysis of piping systems considering the elastic material model and the actual elastic plastic material behaviour is performed. The possibilities of allocating plastic strains to calculated elastic stresses is discussed. A parametric study on straight pipes with the actual elastic plastic material model under pure bending is the basis of deriving the elastic plastic strains for the calculated elastic stresses. Strain limits are suggested which correspond to the different stress categories. The aim is to utilize the deformation possibilities of ductile materials used in German nuclear piping and the allocation of maximum strains to different load categories. Keywords: strain limit, ductile material, stress category. (author)

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

Understanding and coming through PVC-tape-induced stress corrosion cracking in PWR piping system

International Nuclear Information System (INIS)

Shibayama, Motoaki; Shigemoto, Naoya; Noguchi, Shinji; Hirano, Shin-ichi; Takagi, Toshimitsu

2003-01-01

In October 2000, the 24 years old Ikata-1 PWR-type nuclear power plant suffered cracking in pipes of special two lines, where poly vinyl chloride (PVC) tape had been placed and had become baked over time. The existence of residual stress over 100 MPa in the pipes, a bit of chlorine and a feather like-pattern on the crack faces suggested the event was one of stress corrosion cracking. Residual chlorine on the pipes of special two lines was estimated to be 1100 mg/m 2 . A four points bending stress test was performed on the steel plates with the baked on PVC tape in humid air at 80degC. Taking the actual temperature, stress and chlorine on the pipes of the special two lines into consideration, cracking times were estimated to be 12 years and 15 years respectively, which were close to the actual cracking time of 24 years. The authors calculated damage to pipes with fluids of various temperature and duration, and graphed damage contour with a fluid temperature ordinate and a flow duration abscissa. The fluid conditions of major pipes at the Ikata-1 nuclear power plant, which had not received the full inspection, were positioned on so low area on the damage contour that the plant was estimated to be safe for the coming forty years. (author)

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)

About the automated instrumentation in nuclear power plants; Sobre la instrumentacion automatizada en plantas nucleares

Energy Technology Data Exchange (ETDEWEB)

Segovia de los Rios, J. A.; Benitez R, J. S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Garduno G, M P., E-mail: armando.segovia@inin.gob.mx [Instituto Tecnologica de Toluca, Av. Tecnologico s/n, 52140 Metepec, Estado de Mexico (Mexico)

2011-11-15

The automation of the inspection processes and monitoring in nuclear facilities have as main objective the reduction of acquired radiation dose for the operators of the diverse work programs. For example, a typical maintenance task is the problems correction of leaks in the hydraulic facilities of the vapor circuits where is necessary the repair of pipes, measuring and control elements, as the valves. A program of effective maintenance should contemplate strategies of appropriate monitoring for the immediate detection of possible failures, with the purpose of the remedy them opportunely. For this function of failures detection is necessary to have instruments that allow the measuring of the parameters that facilitate their characterization. Given the prevailing conditions in the nuclear facilities, such instruments should possess special characteristics, reason why is necessary a study of them, as well as a careful selection of the susceptible apparatuses of being used. For this reason, this work presents a discussion about some of the existent conditions in the nuclear power plants, as well as the aspects to consider for the automated instrumentation of some places of a nuclear power plant. (Author)

A study on the initiation of condensation-induced water hammer in a long horizontal pipe

International Nuclear Information System (INIS)

Park, Joo Wan

1992-02-01

Condensation-induced water hammer (CIWH) is the most severe and has the highest frequency among the water hammer events occurred in nuclear power plants. Among mechanisms associated with this type of water hammer, the steam and water countercurrent flow in a horizontal pipe is known as the dominant mechanism in Pressurized Water Reactors. The CIWH due to steam-water counter-flow in a long horizontal pipe has been analytically investigated with emphasis on the effect of pipe length, in order to identify the conditions necessary to initiate a water hammer and to provide stability maps describing the zone of water hammer to be avoided with the combination of filling water flowrate and pipe length. Analytical models which can be used to predict the limiting boundaries, upper and lower one, of CIWH initiation have been developed and the calculation results have been compared with the data of an actual incident occurred previously in a nuclear power plant. From the approach used in this study, boundary estimates including simple relationships between critical inlet water flowrates and pipe length-to-diameter on the CIWH initiation in a long horizontal pipe could be made, and several corrective actions to prevent water hammer recurrence could be taken. However, because of the limited understanding of the direct-contact condensation phenomena in the typical range of nuclear power plant operation, it is likely that the overall uncertainty of the analysis results will be high. Therefore, further research on this area including scaling analysis is required

Experimental study on the vibrational characteristics of piping snubbers

International Nuclear Information System (INIS)

Kobatake, K.; Ooka, Y.; Suzuki, M.; Katsuki, T.; Hashimoto, T.

1982-01-01

Oil snubbers have been widely used for the anti-earthquake suports of piping systems in nuclear power plants. Several types of mechanical snubbers are now being considered. Vibration tests were performed on three models to obtain their fundamental characteristics by using a shaking table. From tests on a pendulum structure model, a piping model, and a vessel model, the equivalent stiffness and fundamental characteristics are estimated, and useful suggestions for applications are made

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)

Design of nuclear power generation plants adopting model engineering method

International Nuclear Information System (INIS)

Waki, Masato

1983-01-01

The utilization of model engineering as the method of design has begun about ten years ago in nuclear power generation plants. By this method, the result of design can be confirmed three-dimensionally before actual production, and it is the quick and sure method to meet the various needs in design promptly. The adoption of models aims mainly at the improvement of the quality of design since the high safety is required for nuclear power plants in spite of the complex structure. The layout of nuclear power plants and piping design require the model engineering to arrange rationally enormous quantity of things in a limited period. As the method of model engineering, there are the use of check models and of design models, and recently, the latter method has been mainly taken. The procedure of manufacturing models and engineering is explained. After model engineering has been completed, the model information must be expressed in drawings, and the automation of this process has been attempted by various methods. The computer processing of design is in progress, and its role is explained (CAD system). (Kako, I.)

Mode Identification of Guided Waves in a Curved Pipe