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.)

Piping system damping data at higher frequencies

International Nuclear Information System (INIS)

Ware, A.G.

1987-01-01

Research has been performed at the Idaho National Engineering Laboratory (INEL) for the United States Nuclear Regulatory Commission (USNRC) to determine best-estimate damping values for dynamic analyses of nuclear piping systems excited in the 20 to 100 Hz frequency range. Vibrations in this frequency range are typical of fluid-induced transients, for which no formal pipe damping guidelines exist. The available data found in the open literature and the USNRC/INEL nuclear piping damping data bank were reviewed, and a series of tests on a straight 3-in. (76-mm) piping system and a 5-in. (127-mm) system with several bends and elbows were conducted as part of this research program. These two systems were supported with typical nuclear piping supports that could be changed from test to test during the series. The resulting damping values were ≥ those of the Pressure Vessel Research Committee (PVRC) proposal for unisulated piping. Extending the PVRC damping curve from 20 to 100 Hz at 3% of critical damping would give a satisfactory representation of the test data. This position has been endorsed by the PVRC Technical Committee on Piping Systems. 14 refs

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

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

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

Pipe support optimization in nuclear power plants

International Nuclear Information System (INIS)

Cleveland, A.B.; Kalyanam, N.

1984-01-01

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

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

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.)

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)

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.)

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

International Nuclear Information System (INIS)

Ware, A.G.

1986-01-01

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

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

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

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)

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)

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

Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

International Nuclear Information System (INIS)

Strong, B.R. Jr.; Baschiere, R.J.

1978-01-01

The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the 'force equivalent area' (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques. (Auth.)

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)

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

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

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

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

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

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

International Nuclear Information System (INIS)

Zhong Zhimin; Li Jinsong; Zheng Hui

2012-01-01

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

Nuclear power plant piping damping parametric effects

International Nuclear Information System (INIS)

Ware, A.G.

1983-01-01

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

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

International Nuclear Information System (INIS)

Minematsu, Akiyoshi

1997-01-01

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

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.)

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.

Characterization of radioactive contamination inside pipes with the Pipe Explorer{sup trademark} system

Energy Technology Data Exchange (ETDEWEB)

Cremer, C.D.; Lowry, W.; Cramer, E. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)] [and others

1995-10-01

The U.S. Department of Energy`s nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Historically, this has been attempted using hand held survey instrumentation, surveying only the accessible exterior portions of pipe systems. Difficulty, or inability of measuring threshold surface contamination values, worker exposure, and physical access constraints have limited the effectiveness of this approach. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer{trademark} system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane.

Pipe restraints for nuclear power plants

International Nuclear Information System (INIS)

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

1976-01-01

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

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

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

Pipe support for use in a nuclear system

International Nuclear Information System (INIS)

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

1976-01-01

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

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

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)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Jong, Rudolf Peter de

2004-01-01

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

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

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.

Piping engineering for nuclear power plant

International Nuclear Information System (INIS)

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

1988-01-01

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

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

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

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)

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

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

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

International Nuclear Information System (INIS)

Wu, P.C.

1989-04-01

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

Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

International Nuclear Information System (INIS)

Kendrick, D.T.; Cremer, C.D.; Lowry, W.; Cramer, E.

1995-01-01

The U.S. Department of Energy's nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer trademark system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane. Advantages of this approach include the capability of deploying through constrictions in the pipe, around 90 degrees bends, vertically up and down, and in slippery conditions. Because the detector is transported inside the membrane (which is inexpensive and disposable), it is protected from contamination, which eliminates cross-contamination. Characterization sensors that have been demonstrated with the system thus far include: gamma detectors, beta detectors, video cameras, and pipe locators. Alpha measurement capability is currently under development. A remotely operable Pipe Explorer trademark system has been developed and demonstrated for use in DOE facilities in the decommissioning stage. The system is capable of deployment in pipes as small as 2-inch-diameter and up to 250 feet long. This paper describes the technology and presents measurement results of a field demonstration conducted with the Pipe Explorer trademark system at a DOE site. These measurements identify surface activity levels of U-238 contamination as a function of location in drain lines. Cost savings to the DOE of approximately $1.5 million dollars were realized from this one demonstration

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

Innovative technology summary report: Pipe Explorertrademark system

International Nuclear Information System (INIS)

1996-01-01

The Pipe Explorertrademark system, developed by Science and Engineering Associates, Inc. (SEA), under contract with the US Department of Energy (DOE) Morgantown Energy Technology Center, has been used to transport various characterizing sensors into piping systems that have been radiologically contaminated. DOE's nuclear facility decommissioning program must characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Historically, this has been attempted using hand-held survey instrumentation, surveying only the accessible exterior portions of pipe systems. Various measuring difficulties, and in some cases, the inability to measure threshold surface contamination values and worker exposure, and physical access constraints have limited the effectiveness of traditional survey approaches. The Pipe Explorertrademark system provides a viable alternative. The heart of the system is an air-tight membrane, which is initially spooled inside a canister. The end of the membrane protrudes out of the canister and attaches to the pipe being inspected. The other end of the tubular membrane is attached to the tether and characterization tools. When the canister is pressurized, the membrane inverts and deploys inside the pipe. The characterization detector and its cabling is attached to the tethered end of the membrane. As the membrane is deployed into the pipe, the detector and its cabling is towed into the pipe inside the protective membrane; measurements are taken from within the protective membrane. Once the survey measurements are completed, the process is reversed to retrieve the characterization tools

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.

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

1986-01-01

Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-in. and a pressurized 6-in. diameter carbon steel nuclear pipe systems subjected to high level shaking have been accomplished. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occurred in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate very well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules may be appropriate to cover the ratchet-fatigue failure mode

IEA-R1 renewed primary coolant piping system stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was conducted in 2014. The aim of this work is to perform the stress analysis of the renewed primary piping system of the IEA-R1, taking into account the as built conditions and the pipe modifications. The nuclear research reactor IEA-R1 is a pool type reactor designed by Babcox-Willcox, which is operated by IPEN since 1957. The primary coolant system is responsible for removing the residual heat of the Reactor core. As a part of the life management, a regular inspection detected some degradation in the primary piping system. In consequence, part of the piping system was replaced. The partial renewing of the primary piping system did not imply in major piping layout modifications. However, the stress condition of the piping systems had to be reanalyzed. The structural stress analysis of the primary piping systems is now presented and the final results are discussed. (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

Characterization of pipes, drain lines, and ducts using the pipe explorer system

International Nuclear Information System (INIS)

Cremer, C.D.; Kendrick, D.T.; Cramer, E.

1997-01-01

As DOE dismantles its nuclear processing facilities, site managers must employ the best means of disposing or remediating hundreds of miles of potentially contaminated piping and duct work. Their interiors are difficult to access, and in many cases even the exteriors are inaccessible. Without adequate characterization, it must be assumed that the piping is contaminated, and the disposal cost of buried drain lines can be on the order of $1,200/ft and is often unnecessary as residual contamination levels often are below free release criteria. This paper describes the program to develop a solution to the problem of characterizing radioactive contamination in pipes. The technical approach and results of using the Pipe Explorer trademark system are presented. The heart of the system is SEA's pressurized inverting membrane adapted to transport radiation detectors and other tools into pipes. It offers many benefits over other pipe inspection approaches. It has video and beta/gamma detection capabilities, and the need for alpha detection has been addressed through the development of the Alpha Explorer trademark. These systems have been used during various stages of decontamination and decommissioning of DOE sites, including the ANL CP-5 reactor D ampersand D. Future improvements and extensions of their capabilities are discussed

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

Heat-pipe development for the SPAR space-power system

International Nuclear Information System (INIS)

Ranken, W.A.

1981-01-01

The SPAR space power system design is based on a high temperature fast spectrum nuclear reactor that furnishes heat to a thermoelectric conversion system to generate an electrical power output of 100 kW/sub (e)/. An important feature of this design is the use of alkali metal heat pipes to provide redundant, reliable, and low-loss heat transfer at high temperature. Three sets of heat pipes are used in the system. These include sodium/molybdenum heat pipes to transfer heat from the reactor core to the conversion system, potassium/niobium heat pipes to couple the conversion system to the radiator in a redundant manner, and potassium/titanium heat pipes to distribute rejected heat throughout the radiator surface. The designs of these units are discussed and fabrication methods and testing results are described. 12 figures

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)

Seismic fragility test of a 6-inch diameter pipe system

International Nuclear Information System (INIS)

Chen, W.P.; Onesto, A.T.; DeVita, V.

1987-02-01

This report contains the test results and assessments of seismic fragility tests performed on a 6-inch diameter piping system. The test was funded by the US Nuclear Regulatory Commission (NRC) and conducted by ETEC. The objective of the test was to investigate the ability of a representative nuclear piping system to withstand high level dynamic seismic and other loadings. Levels of loadings achieved during seismic testing were 20 to 30 times larger than normal elastic design evaluations to ASME Level D limits would permit. Based on failure data obtained during seismic and other dynamic testing, it was concluded that nuclear piping systems are inherently able to withstand much larger dynamic seismic loadings than permitted by current design practice criteria or predicted by the probabilistic risk assessment (PRA) methods and several proposed nonlinear methods of failure analysis

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)

Effects of the steam chest on steamhammer analysis for nuclear piping systems

International Nuclear Information System (INIS)

Luk, C.

1975-01-01

When applying the method of characteristics for the steamhammer analysis of a nuclear piping system, if the dynamic fluid behavior in the steam chest is not considered, the boundary condition thus formulated to describe the time-dependent fluid behavior of the steam chest would lead to numerical unstable solution. To overcome this difficulty, the dynamic fluid behavior in the steam chest can be described by a single degree mechanical system. The corresponding flow conditions there are then determined by the time-step amplification method. This dynamic boundary condition reduces the calculated steamhammer loads and helps avoid numerical instability problems in the computing procedure. 4 refs

Fatigue evaluation of socket welded piping in nuclear power plant

International Nuclear Information System (INIS)

Vecchio, R.S.

1996-01-01

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

Nuclear interaction study around beam pipe region in the Tracker system at CMS with 13 TeV data

CERN Document Server

CMS Collaboration

2015-01-01

Analysis is presented to study the material in the Tracker system with nuclear interactions from proton-proton collisions recorded by the CMS experiment at the CERN LHC. The data correspond to an integrated luminosity of 7.3 pb$^{-1}$ at a centre-of-mass energy of 13 TeV collected at 3.8 Tesla magnetic field. With reconstructed nuclear interactions we observe the structure of the material, including beam pipe, in the Tracker system.

Development of automatic pipe welder for nuclear power plant

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Piping data bank and erection system of Angra 2: structure, computational resources and systems

International Nuclear Information System (INIS)

Abud, P.R.; Court, E.G.; Rosette, A.C.

1992-01-01

The Piping Data Bank of Angra 2 called - Erection Management System - Was developed to manage the piping erection of the Nuclear Power Plant of Angra 2. Beyond the erection follow-up of piping and supports, it manages: the piping design, the material procurement, the flow of the fabrication documents, testing of welds and material stocks at the Warehouse. The works developed in the sense of defining the structure of the Data Bank, Computational Resources and System are here described. (author)

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

Applications of equivalent linearization approaches to nonlinear piping systems

International Nuclear Information System (INIS)

Park, Y.; Hofmayer, C.; Chokshi, N.

1997-01-01

The piping systems in nuclear power plants, even with conventional snubber supports, are highly complex nonlinear structures under severe earthquake loadings mainly due to various mechanical gaps in support structures. Some type of nonlinear analysis is necessary to accurately predict the piping responses under earthquake loadings. The application of equivalent linearization approaches (ELA) to seismic analyses of nonlinear piping systems is presented. Two types of ELA's are studied; i.e., one based on the response spectrum method and the other based on the linear random vibration theory. The test results of main steam and feedwater piping systems supported by snubbers and energy absorbers are used to evaluate the numerical accuracy and limitations

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

1987-01-01

Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-inch and a pressurized 6-inch diameter carbon steel nuclear pipe systems subjected to high-level shaking have been accomplished. The high-level shaking loads needed to cause failure were much higher than ASME Code rules would permit with present design limits. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occured in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate reasonably well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules to reduce unneeded conservatisms and to cover the ratchet-fatigue failure mode may be appropriate

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)

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)

Study on seismic design margin based upon inelastic shaking test of the piping and support system

International Nuclear Information System (INIS)

Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

2009-01-01

In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

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

International Nuclear Information System (INIS)

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

2006-07-01

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

Characterization of radioactive contamination inside pipes with the Pipe Explorer{trademark} system. Final report

Energy Technology Data Exchange (ETDEWEB)

Cremer, C.D.; Kendrick, D.T.; Lowry, W.; Cramer, E.

1997-09-30

The Department of Energy (DOE) is currently in the process of decommissioning and dismantling many of its nuclear materials processing facilities that have been in use for several decades. Site managers throughout the DOE complex must employ the safest and most cost effective means to characterize, remediate and recycle or dispose of hundreds of miles of potentially contaminated piping and duct work. The DOE discovered that standard characterization methods were inadequate for its pipes, drains, and ducts because many of the systems are buried or encased. In response to the DOE`s need for a more specialized characterization technique, Science and Engineering Associates, Inc. (SEA) developed the Pipe Explorer{trademark} system through a DOE Office of Science and Technology (OST) contract administered through the Federal Energy Technology Center (FETC). The purpose of this report is to serve as a comprehensive overview of all phases of the Pipe Explorer{trademark} development project. The report is divided into 6 sections. Section 2 of the report provides an overview of the Pipe Explorer{trademark} system, including the operating principles of using an inverting membrane to tow sensors into pipes. The basic components of the characterization system are also described. Descriptions of the various deployment systems are given in Section 3 along with descriptions of the capabilities of the deployment systems. During the course of the development project 7 types of survey instruments were demonstrated with the Pipe Explorer{trademark} and are a part of the basic toolbox of instruments available for use with the system. These survey tools are described in Section 4 along with their typical performance specifications. The 4 demonstrations of the system are described chronologically in Section 5. The report concludes with a summary of the history, status, and future of the Pipe Explorer{trademark} system in Section 6.

Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system. Final report

International Nuclear Information System (INIS)

Cremer, C.D.; Kendrick, D.T.; Lowry, W.; Cramer, E.

1997-01-01

The Department of Energy (DOE) is currently in the process of decommissioning and dismantling many of its nuclear materials processing facilities that have been in use for several decades. Site managers throughout the DOE complex must employ the safest and most cost effective means to characterize, remediate and recycle or dispose of hundreds of miles of potentially contaminated piping and duct work. The DOE discovered that standard characterization methods were inadequate for its pipes, drains, and ducts because many of the systems are buried or encased. In response to the DOE's need for a more specialized characterization technique, Science and Engineering Associates, Inc. (SEA) developed the Pipe Explorer trademark system through a DOE Office of Science and Technology (OST) contract administered through the Federal Energy Technology Center (FETC). The purpose of this report is to serve as a comprehensive overview of all phases of the Pipe Explorer trademark development project. The report is divided into 6 sections. Section 2 of the report provides an overview of the Pipe Explorer trademark system, including the operating principles of using an inverting membrane to tow sensors into pipes. The basic components of the characterization system are also described. Descriptions of the various deployment systems are given in Section 3 along with descriptions of the capabilities of the deployment systems. During the course of the development project 7 types of survey instruments were demonstrated with the Pipe Explorer trademark and are a part of the basic toolbox of instruments available for use with the system. These survey tools are described in Section 4 along with their typical performance specifications. The 4 demonstrations of the system are described chronologically in Section 5. The report concludes with a summary of the history, status, and future of the Pipe Explorer trademark system in Section 6

Experience with the TUeV pipe monitoring system at the Grohnde nuclear power station

International Nuclear Information System (INIS)

Dittmar, H.; Hofstoetter, P.

1995-01-01

A special pipe monitoring system has been developed by TUeV Rheinland during the construction, commissioning and operation of the Grohnde nuclear power station. On the basis of measurements during construction and commissioning a basic monitoring system has been developed, using not only a system of sophisticated sensors that had been permanently installed from the beginning but also a large number of quite simple additional sensors. Measurements were taken before, during and after inspections and led to the discovery of unexpected and high stresses during service as well as to long-term changes over a period of years.Special measurements were taken with high temperature strain gauges and thermocouples to identify problems such as temperature layering. A special on-line measuring device was developed and used for the continuous monitoring of temperatures during operation.All these measurements help to identify out areas with high stresses or service conditions giving rise to high loads, in order on the one hand to prevent damage and on the other hand to prove that the pipes are functioning within their design parameters without problems. ((orig.))

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

Seismic design of piping systems

International Nuclear Information System (INIS)

Anglaret, G.; Beguin, J.L.

1986-01-01

This paper deals with the method used in France for the PWR nuclear plants to derive locations and types of supports of auxiliary and secondary piping systems taking earthquake in account. The successive steps of design are described, then the seismic computation method and its particular conditions of applications for piping are presented. The different types of support (and especially seismic ones) are described and also their conditions of installation. The method used to compare functional tests results and computation results in order to control models is mentioned. Some experiments realised on site or in laboratory, in order to validate models and methods, are presented [fr

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)

Acoustic analysis of a piping system

International Nuclear Information System (INIS)

Misra, A.S.; Vijay, D.K.

1996-01-01

Acoustic pulsations in the Darlington Nuclear Generating Station, a 881 MW CANDU, primary heat transport piping system caused fuel bundle failures under short term operations. The problem was successfully analyzed using the steady-state acoustic analysis capability of the ABAQUS program. This paper describes in general, modelling of low amplitude acoustic pulsations in a liquid filled piping system using ABAQUS. The paper gives techniques for estimating the acoustic medium properties--bulk modulus, fluid density and acoustic damping--and modelling fluid-structure interactions at orifices and elbows. The formulations and techniques developed are benchmarked against the experiments given in 3 cited references. The benchmark analysis shows that the ABAQUS results are in excellent agreement with the experiments

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

Automatic seismic support design of piping system by an object oriented expert system

International Nuclear Information System (INIS)

Nakatogawa, T.; Takayama, Y.; Hayashi, Y.; Fukuda, T.; Yamamoto, Y.; Haruna, T.

1990-01-01

The seismic support design of piping systems of nuclear power plants requires many experienced engineers and plenty of man-hours, because the seismic design conditions are very severe, the bulk volume of the piping systems is hyge and the design procedures are very complicated. Therefore we have developed a piping seismic design expert system, which utilizes the piping design data base of a 3 dimensional CAD system and automatically determines the piping support locations and support styles. The data base of this system contains the maximum allowable seismic support span lengths for straight piping and the span length reduction factors for bends, branches, concentrated masses in the piping, and so forth. The system automatically produces the support design according to the design knowledge extracted and collected from expert design engineers, and using design information such as piping specifications which give diameters and thickness and piping geometric configurations. The automatic seismic support design provided by this expert system achieves in the reduction of design man-hours, improvement of design quality, verification of design result, optimization of support locations and prevention of input duplication. In the development of this system, we had to derive the design logic from expert design engineers and this could not be simply expressed descriptively. Also we had to make programs for different kinds of design knowledge. For these reasons we adopted the object oriented programming paradigm (Smalltalk-80) which is suitable for combining programs and carrying out the design work

Refurbishment of the IEAR1 primary coolant system piping supports

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

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

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)

Flow induced vibrations in a PWR piping system

International Nuclear Information System (INIS)

Seligmann, D.C.; Guillou, J.P.

1995-01-01

In this paper, we present and industrial study of the dynamic behaviour of the piping system of a French 1300 M We nuclear power plant. High-amplitude vibrations had been noticed on a safeguard system during the periodical operation startup tests. These vibrations, due to acoustical pump sources, cause fatigue-damage and it is therefore necessary to propose an estimation of the service-life of the piping and to propose modification of piping system to reduce vibrations. First, we define a mechanical model readjusted according to gauged vibratory speeds and construct a vibro-acoustic coupled model and a pump-behaviour model as a source of excitation. Second, we simulate a modification of the supports. The influence of this modification is analysed by comparison of the root mean square values of vibratory speeds and the stresses between the initial system and the modified system. 3 refs., 7 figs

IPIRG-2 task 1 - pipe system experiments with circumferential cracks in straight-pipe locations. Final report, September 1991--November 1995

Energy Technology Data Exchange (ETDEWEB)

Scott, P.; Olson, R.; Marschall, C.; Rudland, D. [and others

1997-02-01

This report presents the results from Task 1 of the Second International Piping Integrity Research Group (IPIRG-2) program. The IPIRG-2 program is an international group program managed by the US Nuclear Regulatory Commission (US NRC) and funded by a consortium of organizations from 15 nations including: Bulgaria, Canada, Czech Republic, France, Hungary, Italy, Japan, Republic of Korea, Lithuania, Republic of China, Slovak Republic, Sweden, Switzerland, the United Kingdom, and the United States. The objective of the program was to build on the results of the IPIRG-1 and other related programs by extending the state-of-the-art in pipe fracture technology through the development of data needed to verify engineering methods for assessing the integrity of nuclear power plant piping systems that contain defects. The IPIRG-2 program included five main tasks: Task 1 - Pipe System Experiments with Flaws in Straight Pipe and Welds Task 2 - Fracture of Flawed Fittings Task 3 - Cyclic and Dynamic Load Effects on Fracture Toughness Task 4 - Resolution of Issues From IPIRG-1 and Related Programs Task 5 - Information Exchange Seminars and Workshops, and Program Management. The scope of this report is to present the results from the experiments and analyses associated with Task 1 (Pipe System Experiments with Flaws in Straight Pipe and Welds). The rationale and objectives of this task are discussed after a brief review of experimental data which existed after the IPIRG-1 program.

IPIRG-2 task 1 - pipe system experiments with circumferential cracks in straight-pipe locations. Final report, September 1991--November 1995

International Nuclear Information System (INIS)

Scott, P.; Olson, R.; Marschall, C.; Rudland, D.

1997-02-01

This report presents the results from Task 1 of the Second International Piping Integrity Research Group (IPIRG-2) program. The IPIRG-2 program is an international group program managed by the US Nuclear Regulatory Commission (US NRC) and funded by a consortium of organizations from 15 nations including: Bulgaria, Canada, Czech Republic, France, Hungary, Italy, Japan, Republic of Korea, Lithuania, Republic of China, Slovak Republic, Sweden, Switzerland, the United Kingdom, and the United States. The objective of the program was to build on the results of the IPIRG-1 and other related programs by extending the state-of-the-art in pipe fracture technology through the development of data needed to verify engineering methods for assessing the integrity of nuclear power plant piping systems that contain defects. The IPIRG-2 program included five main tasks: Task 1 - Pipe System Experiments with Flaws in Straight Pipe and Welds Task 2 - Fracture of Flawed Fittings Task 3 - Cyclic and Dynamic Load Effects on Fracture Toughness Task 4 - Resolution of Issues From IPIRG-1 and Related Programs Task 5 - Information Exchange Seminars and Workshops, and Program Management. The scope of this report is to present the results from the experiments and analyses associated with Task 1 (Pipe System Experiments with Flaws in Straight Pipe and Welds). The rationale and objectives of this task are discussed after a brief review of experimental data which existed after the IPIRG-1 program

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

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

Pipe support

International Nuclear Information System (INIS)

Pollono, L.P.

1979-01-01

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

Seismic analysis of piping systems subjected to multiple support excitations

International Nuclear Information System (INIS)

Sundararajan, C.; Vaish, A.K.; Slagis, G.C.

1981-01-01

The paper presents the results of a comparative study between the multiple response spectrum method and the time-history method for the seismic analysis of nuclear piping systems subjected to different excitation at different supports or support groups. First, the necessary equations for the above analysis procedures are derived. Then, three actual nuclear piping systems subjected to single and multiple excitations are analyzed by the different methods, and extensive comparisons of the results (stresses) are made. Based on the results, it is concluded that the multiple response spectrum analysis gives acceptable results as compared to the ''exact'', but much more costly, time-history analysis. 6 refs

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').

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)

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

Directory of Open Access Journals (Sweden)

Jai Hak Park

2015-04-01

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

Manufacture of piping components for nuclear power plants

International Nuclear Information System (INIS)

Bartecek, R.

1983-01-01

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

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

Flow induced vibrations in a PWR piping system

International Nuclear Information System (INIS)

Seligmann, D.; Guillou, J.

1995-11-01

During a recurring bench test of an operating system, high amplitude vibrations have been observed on a safety piping system of a nuclear power plant. Due to the source of the pumps, these vibrations lead to wear damage and it is therefore necessary to estimate the life time of the piping system. This paper describes the methodology used to study the dynamic behaviour and to analyze the damage of a piping system submitted to internal flow. Starting from an experimental modal analysis of the piping system when not i service, we analyse the main parameters of the mechanical behaviour. Following this analysis, we obtain a mechanical model fitting the first experimental modes. On this basis, we build a vibro-acoustical model. This model takes into account the influence of the acoustical pipe length, both above and below the mechanical part, the modelling of acoustical components, the speed of sound. We did not experimentally characterize the pumps. Therefore, we use a numerical model in order to simulate the behaviour of the pumps. This model is based on the theory of the transfer matrix and takes into account the geometric and the hydraulic characteristics of the pump.The modelling of both sources (suction and discharge) connected to the pump is formed by contributions from a source corresponding to the turbulent noise at low frequency, a source at blade passage frequency. This model has been experimentally validated in a laboratory. The final results of the modelling of the complete piping system are in a complete accord with experimental measurements. (author). 3 refs., 7 figs

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

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

Damping in LMFBR pipe systems

International Nuclear Information System (INIS)

Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

1983-06-01

LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

Piping reliability model development, validation and its applications to light water reactor piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-01-01

A brief description is provided of a three-year effort undertaken by the Lawrence Livermore National Laboratory for the piping reliability project. The ultimate goal of this project is to provide guidance for nuclear piping design so that high-reliability piping systems can be built. Based on the results studied so far, it is concluded that the reliability approach can undoubtedly help in understanding not only how to assess and improve the safety of the piping systems but also how to design more reliable piping systems

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

Pipe damping

International Nuclear Information System (INIS)

Ware, A.G.

1985-01-01

Studies are being conducted at the Idaho National Engineering Laboratory to determine whether an increase in the damping values used in seismic structural analyses of nuclear piping systems is justified. Increasing the allowable damping would allow fewer piping supports which could lead to safer, more reliable, and less costly piping systems. Test data from availble literature were examined to determine the important parameters contributing to piping system damping, and each was investigated in separate-effects tests. From the combined results a world pipe damping data bank was established and multiple regression analyses performed to assess the relative contributions of the various parameters. The program is being extended to determine damping applicable to higher frequency (33 to 100 Hz) fluid-induced loadings. The goals of the program are to establish a methodology for predicting piping system damping and to recommend revised guidelines for the damping values to be included in analyses

Design evaluation on sodium piping system and comparison of the design codes

International Nuclear Information System (INIS)

Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

2015-01-01

A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

Design evaluation on sodium piping system and comparison of the design codes

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

2015-03-15

A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

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-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

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.

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)

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.

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)

In-situ rehabilitation cleans, lines, and renews pipe systems

International Nuclear Information System (INIS)

Munden, B.A.

1990-01-01

This article discusses how, in the past five years, developments in coating and lining material technology have found their way into pipe line application and have yielded successful results. The thick film, high solids material often used to repair tanks, vessels and offshore structures has now been adapted for existing pipe lines. One of the most promising of these systems in successful service is an epoxy, high solids (95%) material originally developed for nuclear service as a lining for reactor containment vessels

Evaluation of temporary non-code repairs in safety class 3 piping systems

International Nuclear Information System (INIS)

Godha, P.C.; Kupinski, M.; Azevedo, N.F.

1996-01-01

Temporary non-ASME Code repairs in safety class 3 pipe and piping components are permissible during plant operation in accordance with Nuclear Regulatory Commission Generic Letter 90-05. However, regulatory acceptance of such repairs requires the licensee to undertake several timely actions. Consistent with the requirements of GL 90-05, this paper presents an overview of the detailed evaluation and relief request process. The technical criteria encompasses both ductile and brittle piping materials. It also lists appropriate evaluation methods that a utility engineer can select to perform a structural integrity assessment for design basis loading conditions to support the use of temporary non-Code repair for degraded piping components. Most use of temporary non-code repairs at a nuclear generating station is in the service water system which is an essential safety related system providing the ultimate heat sink for various plant systems. Depending on the plant siting, the service water system may use fresh water or salt water as the cooling medium. Various degradation mechanisms including general corrosion, erosion/corrosion, pitting, microbiological corrosion, galvanic corrosion, under-deposit corrosion or a combination thereof continually challenge the pressure boundary structural integrity. A good source for description of corrosion degradation in cooling water systems is provided in a cited reference

Relative conservatisms of combination methods used in response spectrum analyses of nuclear piping systems

International Nuclear Information System (INIS)

Gupta, S.; Kustu, O.; Jhaveri, D.P.; Blume, J.A.

1983-01-01

The paper presents the conclusions of a comprehensive study that investigated the relative conservatisms represented by various combination techniques. Two approaches were taken for the study, producing mutually consistent results. In the first, 20 representative nuclear piping systems were systematically analyzed using the response spectrum method. The total response was obtained using nine different combination methods. One procedure, using the SRSS method for combining spatial components of response and the 10% method for combining the responses of different modes (which is currently acceptable to the U.S. NRC), was the standard for comparison. Responses computed by the other methods were normalized to this standard method. These response ratios were then used to develop cumulative frequency-distribution curves, which were used to establish the relative conservatism of the methods in a probabilistic sense. In the second approach, 30 single-degree-of-freedom (SDOF) systems that represent different modes of hypothetical piping systems and have natural frequencies varying from 1 Hz to 30 Hz, were analyzed for 276 sets of three-component recorded ground motion. A set of hypothetical systems assuming a variety of modes and frequency ranges was developed. The responses of these systems were computed from the responses of the SDOF systems by combining the spatial response components by algebraic summation and the individual mode responses by the Navy method, or combining both spatial and modal response components using the SRSS method. Probability density functions and cumulative distribution functions were developed for the ratio of the responses obtained by both methods. (orig./HP)

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

Investigation on method of elasto-plastic analysis for piping system (benchmark analysis)

International Nuclear Information System (INIS)

Kabaya, Takuro; Kojima, Nobuyuki; Arai, Masashi

2015-01-01

This paper provides method of an elasto-plastic analysis for practical seismic design of nuclear piping system. JSME started up the task to establish method of an elasto-plastic analysis for nuclear piping system. The benchmark analyses have been performed in the task to investigate on method of an elasto-plastic analysis. And our company has participated in the benchmark analyses. As a result, we have settled on the method which simulates the result of piping exciting test accurately. Therefore the recommended method of an elasto-plastic analysis is shown as follows; 1) An elasto-plastic analysis is composed of dynamic analysis of piping system modeled by using beam elements and static analysis of deformed elbow modeled by using shell elements. 2) Bi-linear is applied as an elasto-plastic property. Yield point is standardized yield point multiplied by 1.2 times, and second gradient is 1/100 young's modulus. Kinematic hardening is used as a hardening rule. 3) The fatigue life is evaluated on strain ranges obtained by elasto-plastic analysis, by using the rain flow method and the fatigue curve of previous studies. (author)

Calculation of the pipes failure probability of the Rcic system of a nuclear power station by means of software WinPRAISE 07

International Nuclear Information System (INIS)

Jasso G, J.; Diaz S, A.; Mendoza G, G.; Sainz M, E.; Garcia de la C, F. M.

2014-10-01

The growth and the cracks propagation by fatigue are a typical degradation mechanism that is presented in the nuclear industry as in the conventional industry; the unstable propagation of a crack can cause the catastrophic failure of a metallic component even with high ductility; for this reason, activities of programmed maintenance have been established in the industry using inspection and visual techniques and/or ultrasound with an established periodicity allowing to follow up to these growths, controlling the undesirable effects; however, these activities increase the operation costs; and in the peculiar case of the nuclear industry, they increase the radiation exposure to the participant personnel. The use of mathematical processes that integrate concepts of uncertainty, material properties and the probability associated to the inspection results, has been constituted as a powerful tool of evaluation of the component reliability, reducing costs and exposure levels. In this work the evaluation of the failure probability by cracks growth preexisting by fatigue is presented, in pipes of a Reactor Core Isolation Cooling system (Rcic) in a nuclear power station. The software WinPRAISE 07 (Piping Reliability Analysis Including Seismic Events) was used supported in the probabilistic fracture mechanics principles. The obtained values of failure probability evidenced a good behavior of the analyzed pipes with a maximum order of 1.0 E-6, therefore is concluded that the performance of the lines of these pipes is reliable even extrapolating the calculations at 10, 20, 30 and 40 years of service. (Author)

Analysis of pipe stress using CAESAR II code

International Nuclear Information System (INIS)

Sitandung, Y.B.; Bandriyana, B.

2002-01-01

Analysis of this piping stress with the purpose of knowing stress distribution piping system in order to determine pipe supports configuration. As an example of analysis, Gas Exchanger to Warm Separator Line was chosen with, input data was firstly prepared in a document, i.e. piping analysis specification that its content named as pipe characteristics, material properties, operation conditions, guide equipment's and so on. Analysis result such as stress, load, displacement and the use support type were verified based on requirements in the code, standard, and regularities were suitable with piping system condition analyzed. As the proof that piping system is in safety condition, it can be indicated from analysis results (actual loads) which still under allowable load. From the analysis steps that have been done CAESAR II code fulfill requirements to be used as a tool of piping stress analysis as well as nuclear and non nuclear installation piping system

INEL/USNRC pipe damping experiments and studies

International Nuclear Information System (INIS)

Ware, A.G.

1987-08-01

Since the previous paper on this subject presented at the 8th SMiRT Conference, the Idaho National Engineering Laboratory (INEL) has conducted further research on piping system damping for the United States Nuclear Regulatory Commission (USNRC). These efforts have included vibration tests on two laboratory piping systems at response frequencies up to 100 Hz, and damping data calculations from both of these two systems and from a third laboratory piping system test series. In addition, a statistical analysis was performed on piping system damping data from tests representative of seismic and hydrodynamic events of greater than minimal excitation. The results of this program will be used to assist regulators in establishing suitable damping values for use in dynamic analyses of nuclear piping systems, and in revising USNRC Regulatory Guide (RG) 1.61

Water Hammer Mitigation on Postulated Pipe Break of Feed Water System

International Nuclear Information System (INIS)

Seong, Ho Je; Woo, Kab Koo; Cho, Keon Taek

2008-01-01

The Feed Water (FW) system supplies feedwater from the deaerator storage tank to the Steam Generators(S/G) at the required pressure, temperature, flow rate, and water chemistry. The part of FW system, from the S/G to Main Steam Valve House just outside the containment building wall, is designed as safety grade because of its safety function. According to design code the safety related system shall be designed to protect against dynamic effects that may results from a pipe break on high energy lines such as FW system. And the FW system should be designed to minimize blowdown volume of S/G secondary side during the postulated pipe break. Also the FW system should be designed to prevent the initiation or to minimize the effects of water hammer transients which may be induced by the pipe break. This paper shows the results of the hydrodynamic loads induced by the pipe break and the optimized design parameters to mitigate water hammer loads of FW system for Shin-Kori Nuclear Power Plant Unit 3 and 4 (SKN 3 and 4)

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

A leak-before-break strategy for CANDU primary piping systems

International Nuclear Information System (INIS)

Aggarwal, M.L.; Kozluk, M.J.; Lin, T.C.; Manning, B.W.; Vijay, D.K.

1986-01-01

Recent advances in elastic-plastic fracture mechanics have made it possible to assess the stability of cracks in ductile piping systems. These technological developments have been used by Ontario Hydro as the nucleus of an approach for demonstrating that CANDU primary heat transport piping systems will not break catastrophically; at worst they would leak at a detectable rate. This leak-before-break approach has been taken on the Darlington nuclear generating station as a design stage alternative to the provision of pipe whip restraints on large diameter, primary heat transport system piping. Positive conclusions reached via this approach are considered sufficient to exclude the requirement to provide protective devices, such as pipe whip restraints. In arriving at the proposed leak-before-break approach a review of current and proposed leak-before-break licensing positions of other jurisdictions (particularly those in the United States and the Federal Republic of Germany) was carried out. The approach presented makes use of recent American developments in the area of elastic-plastic fracture mechanics. It also gives consideration to aspects which are unique to the pressurized heavy water (CANDU) reactors used by Ontario Hydro. The proposed leak-before-break approach is described and its use is illustrated by applying it to the Darlington generating station primary heat transport system pump suction piping. (author)

Development of an evaluation method for seismic isolation systems of nuclear power facilities. Development of crossover piping design method for seismic isolation systems

International Nuclear Information System (INIS)

Otoyo, Teruyoshi; Otani, Akihito; Otani, Akihito; Fukushima, Shunsuke; Jimbo, Masakazu; Yamamoto, Tomofumi; Sakakida, Takaaki; Onishi, Shigenobu

2014-01-01

In the conceptual design of seismic isolation systems of nuclear power facilities, there exist two types of installation. The first type is to isolate both the reactor and the turbine buildings, the other is to isolate only the reactor building. In the latter type, the crossover piping, which installed between the isolated and the non-isolated buildings, is excited and deformed by the different motions of those buildings. In this study, shaking tests of 1/10 scaled model of the main steam piping and FEM analyses under multiple support excitation conditions have been performed to investigate the vibration behavior of the crossover piping. It was confirmed that modal time-history analyses could be in good agreement with the shaking test results. Also, Numerous combination methods were investigated by comparing response spectrum analyses and modal time-history analyses. In conclusion, response spectrum analyses using SRSS combinations could correspond to time-history analyses. (author)

Vibration analysis of the piping system using the modal analysis method, 1

International Nuclear Information System (INIS)

Fujikawa, Takeshi; Kurohashi, Michiya; Inoue, Yoshio

1975-01-01

Modal analysis method was developed for the vibration analysis of piping system in nuclear or chemical plants, with finite element theory, and verified by sinusoidal vibration method. The natural vibration equation for pipings was derived with stiffness, attenuation and mass matrices, and eigenvalues are obtained with usual method, then the forced vibration equation for pipings was derived with the same manner, and the special solutions are given by modal method from the eigenvalues of the natural vibration equation. Three simple piping models (one, two and three dimensional) were made, and the natural vibration frequency was measured with forced input from an electrical dynamic shaker and a sound speaker. The experimental values of natural vibration frequency showed good agreement with the results by the analytical method. Therefore the theoretical approach for piping system vibration was proved to be valid. (Iwase, T.)

The intermittent contact impact problem in piping systems of nuclear reactor

International Nuclear Information System (INIS)

Martin, A.; Ricard, A.; Millard, A.

1981-09-01

The intermittent contact problem is important in many pipe whip studies, specially as to the safety of nuclear reactors. The impact concept adopted is that of instantaneous impact, so that at the time of impact the two impacting structures instantaneously acquire the same velocity in the impact direction. Energy is dissipated by some mechanism whose spatial and temporal scale is small compared to these scales in the discrete model. This dissipation is associated with local plastic deformation. Different solutions are presented for solving this problem. The first one is a generalization of the modal superposition method, when the nonlinearities of the structure are only due to impact between structural components; the other ones are included in a step by step time history and can take in account geometrical non linearities and of behavior. Some industrial applications in nuclear technology are presented

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.)

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

Nuclear power plant piping prefabrication and assembly

International Nuclear Information System (INIS)

Schmidt, H.

1990-01-01

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

Engineering design aspects of the heat-pipe power system

Science.gov (United States)

Capell, B. M.; Houts, M. G.; Poston, D. I.; Berte, M.

1997-01-01

The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.

Engineering design aspects of the heat-pipe power system

International Nuclear Information System (INIS)

Capell, B.M.; Houts, M.G.; Poston, D.I.; Berte, M.

1997-10-01

The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations

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

Design and Integrity Evaluation of High-temperature Piping Systems in the STELLA-2 Sodium Test Facility

Energy Technology Data Exchange (ETDEWEB)

Son, Seok-Kwon; Lee, Hyeong-Yeon; Eoh, JaeHyuk; Kim, Jong-Bum; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ju, Yong-Sun [KOASIS Inc., Daejeon (Korea, Republic of)

2016-09-15

In this study, elevated temperature design and integrity evaluation have been conducted using two different piping design codes for the high-temperature piping systems of sodium integral effect test loop for safety simulation and assessment(STELLA-2) being developed by KAERI(Korea Atomic Energy Research Institute). The design code of ASME B31.1 for power piping and French nuclear grade piping design guideline, RCC-MRx RD-3600 were applied, and conservatism of those codes was quantified based on the piping integrity evaluation results. The piping system of Model DHRS, Model IHTS and PSLS are to be installed in STELLA-2. The integrity evaluation results for the three piping systems according to the two design codes showed that integrity of the piping system was confirmed. As a code comparison result, ASME B31.1 was shown to be more conservative for sustained loads while RD-3600 was more conservative for thermal loads compared to B31.1.

Seismic margins and calibration of piping systems

International Nuclear Information System (INIS)

Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

1985-01-01

The Seismic Safety Margins Research Program (SSMRP) is a US Nuclear Regulatory Commission-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its objective is to develop a complete, fully coupled analysis procedure for estimating the risk of earthquake-induced radioactive release from a commercial nuclear power plant and to determine major contributors to the state-of-the-art seismic and systems analysis process and explicitly includes the uncertainties in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I of SSMRP, the overall seismic risk assessment methodology was developed and assembled. The application of this methodology to the seismic PRA (Probabilistic Risk Assessment) at the Zion Nuclear Power Plant has been documented. This report documents the method deriving response factors. The response factors, which relate design calculated responses to best estimate values, were used in the seismic response determination of piping systems for a simplified seismic probablistic risk assessment. 13 references, 31 figures, 25 tables

Efficient erection of a piping unit in a nuclear power station

International Nuclear Information System (INIS)

Halstrick, V.; Peters, G.

1986-01-01

In consideration of the negative experience gathered in the past extensive project logistics are required for the erection of piping units in a nuclear power station in order to be able to recognize and master the numerous influences and different marginal conditions with reasonable certainty and at an early stage. The utilization of requirements from the analysis of experience for the conception of project management begins with the erection planning and results in check lists for the execution of erection. During production planning these check lists are verified for realization. Because of the extensive data, EDP-aided systems are applied for checking and controlling the flow of information and material. A dialogue-aided system is presented for project planning and controlling which enables a transparent and farsighted execution of a project. By means of comparable piping units it is demonstrated that due to the created controlling system a great success becomes obvious in relation to the past. (orig.) [de

Integrity assessment of the cold leg piping system in a PWR

International Nuclear Information System (INIS)

Mayfield, M.E.; Leis, B.N.

1981-01-01

The purpose of this paper is to examine the integrity of a nuclear piping system, designed in accordance with Section III, in the context of a damage tolerance analysis procedure. Such a procedure directly addresses the defects and cyclic loadings that are responsible for the above noted exceptions. The analysis and results reported here are for a fatigue life analysis of the Cold Leg piping in a PWR. This piping system is particularly important from a safety standpoint since a large break is a possible initiator of a core meltdown accident. The analysis employs LEFM concepts to determine the time between the initial start-up and (1) formation of a leak, (2) detection of the leak, and (3) the final fracture of the piping. Both longitudinal and circumferential defects are considered. The defects are assumed to propagate from the pipe I.D. in a self-similar manner. Inputs to the analysis were derived from information supplied by plant operators and vendors, published data, and 'expert opinions'. The life was computed using a linear damage accumulation. (orig./GL)

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)

Role of theoretical dynamics in vibration diagnostics of pipe systems

International Nuclear Information System (INIS)

Rejent, B.

1992-01-01

The importance of vibration diagnostics of pipe systems and the relevance of theoretical dynamics are shown using examples. The problems are discussed of vibration diagnostics of the primary circuit of a nuclear power plant with viscous seismic dampers installed. (M.D.) 7 figs., 5 refs

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

Alternative methods for the seismic analysis of piping systems

International Nuclear Information System (INIS)

Anon.

1993-01-01

This document is a review of 12 methods and criteria for the seismic analysis of piping systems. Each of the twelve chapters in this document cover the important technical aspects of a given method. The technical aspects presented are those the Subcommittee on Dynamic Stress Criteria believe important to the application of the method, and should not be considered as a positive or negative endorsement for any of the methods. There are many variables in an analysis of a piping system that can influence the selection of the analysis method and criteria to be applied. These variable include system configuration, technical issues, precedent, licensing considerations, and regulatory acceptance. They must all be considered in selecting the appropriate seismic analysis method and criteria. This is relevant for nuclear power plants

Reactor Primary Coolant System Pipe Rupture Study. Progress report No. 32, July--December 1974

International Nuclear Information System (INIS)

1975-03-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue studies focused on Elastic/Plastic ASME Code Design Rules, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, and (c) studies directed at quantifying weld sensitization in T-304 stainless steel. (auth)

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.)

Reliability of piping system components. Volume 4: The pipe failure event database

Energy Technology Data Exchange (ETDEWEB)

Nyman, R; Erixon, S [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B [RSA Technologies, Visat, CA (United States)

1996-07-01

Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. 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 failure. 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. 42 refs, 25 figs.

Reliability of piping system components. Volume 4: The pipe failure event database

International Nuclear Information System (INIS)

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

1996-07-01

Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. 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 failure. 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. 42 refs, 25 figs

IEA-R1 primary and secondary coolant piping systems coupled stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A.; Mattar Neto, Miguel

2013-01-01

The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The operation to 5 MW power limit was only possible after the conduction of life management and modernization programs in the last two decades. In these programs the components of the coolant systems, which are responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping systems is presented and the final results are discussed. (author)

Evaluation of LBB margin of nuclear piping systems

International Nuclear Information System (INIS)

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

1999-04-01

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

Evaluation of LBB margin of nuclear piping systems

Energy Technology Data Exchange (ETDEWEB)

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

1999-04-15

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

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

International Nuclear Information System (INIS)

Mc Donald, B.N.

1976-01-01

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

46 CFR 153.280 - Piping system design.

Science.gov (United States)

2010-10-01

... 46 Shipping 5 2010-10-01 2010-10-01 false Piping system design. 153.280 Section 153.280 Shipping... BULK LIQUID, LIQUEFIED GAS, OR COMPRESSED GAS HAZARDOUS MATERIALS Design and Equipment Piping Systems and Cargo Handling Equipment § 153.280 Piping system design. (a) Each cargo piping system must meet...

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).

Reliability of piping system components. Framework for estimating failure parameters from service data

International Nuclear Information System (INIS)

Nyman, R.; Hegedus, D.; Tomic, B.; Lydell, B.

1997-12-01

This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed 'PFCA'-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies

Reliability of piping system components. Framework for estimating failure parameters from service data

Energy Technology Data Exchange (ETDEWEB)

Nyman, R [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hegedus, D; Tomic, B [ENCONET Consulting GesmbH, Vienna (Austria); Lydell, B [RSA Technologies, Vista, CA (United States)

1997-12-01

This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed `PFCA`-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies. 63 refs, 30 tabs, 22 figs.

Guidelines and criteria for nuclear piping and support evaluation and design

International Nuclear Information System (INIS)

Rehn, D.L.; Stout, D.H. Jr.; Minichiello, J.C.

1993-05-01

The EPRI Research Project 2967-2 has set its fundamental goal to be the development of realistic guidelines and criteria for piping and pipe support design and evaluation. The focus is on items that are most critical to utilities and consists of a variety of tasks relating to piping and pipe support design. One objective of this report is to summarize the recommendations from the seven task reports of the first phase of the project and to provide examples of how to use those recommendations. Criteria and methods for evaluating both short and long term system operation are addressed. Benefits gained from applying the recommendations to actual systems are discussed. The report also reviews other work currently being done within the nuclear industry and assesses the impact of that work on the recommended criteria/methods of this project. The second objective of the report is to discuss possible changes needed in the governing codes or licensing commitments in order to implement the recommendations. Finally, the report describes further research which can be done to advance the criteria presented and answer questions concerning applicability of the proposed criteria to designs not tested/investigated. The basic conclusion reached in the project is that many of the criteria/methods used today in piping analysis/design are overly conservative. The report's conclusion is supported by extensive literature searches, tests, and analyses. The report presents a robust source of reference to utilities which wish to implement changes in criteria and methods. Most of the criteria and methodologies described in the seven task reports and summarized in the following sections will require some effort in licensing or Code changes

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

Acoustic system for pipe rupture monitoring and leak detection

International Nuclear Information System (INIS)

Herzog, W.; Jonas, H.

1982-06-01

As a safety aspect pipe rupture and leakage effects are of particular interest in nuclear power plants where severe consequences for the reactor may result. Counter measures against postulated pipe breaks and leakages in nuclear power plants are necessary whenever the main safety goals: safe shut-down, safe afterheat removal and retention of radioactivity, are endangered. The requirements to be met by a leak detection system depend on the time available for counter actions. If this time is short so that automatic actions are necessary the German safety criteria for nuclear power plants (Criterion 6.1) require two physically diverse signals to be monitored. One fairly obvious possibility of leak detection is to monitor process parameters (pressure, flow). As a diverse signal physical parameters outside the process may be employed: pressure transients temperature, humidity are principally suitable. In practical application, however, it is difficult to predict these parameters by way of calculation in order to establish the required set-point of the monitoring system. Experimental determination is possible only in special cases. A study of several ways of diverse leak detection methods leads to the very promising acoustic method. We investigated experimentally the feasibility of monitoring the sound created by a leakage. Air borne sound as well as body borne sound was analyzed

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

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

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)

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.)

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

Energy Technology Data Exchange (ETDEWEB)

Mohan, R.; Marshall, C.; Ghadiali, N.; Wilkowski, G. [Battelle, Columbus, OH (United States)

1997-04-01

This paper summarizes work on angled through-wall-crack initiation and combined loading effects on ferritic nuclear pipe performed as part of the Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks In Piping an Piping Welds{close_quotes}. The reader is referred to Reference 1 for details of the experiments and analyses conducted as part of this program. The major impetus for this work stemmed from the observation that initially circumferentially oriented cracks in carbon steel pipes exhibited a high tendency to grow at a different angle when the cracked pipes were subjected to bending or bending plus pressure loads. This failure mode was little understood, and the effect of angled crack grown from an initially circumferential crack raised questions about how cracks in a piping system subjected to combined loading with torsional stresses would behave. There were three major efforts undertaken in this study. The first involved a literature review to assess the causes of toughness anisotropy in ferritic pipes and to develop strength and toughness data as a function of angle from the circumferential plane. The second effort was an attempt to develop a screening criterion based on toughness anisotropy and to compare this screening criterion with experimental pipe fracture data. The third and more significant effort involved finite element analyses to examine why cracks grow at an angle and what is the effect of combined loads with torsional stresses on a circumferentially cracked pipe. These three efforts are summarized.

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

Application of new developments in coupled seismic analysis of piping systems

International Nuclear Information System (INIS)

Gupta, A.; Gupta, A.K.

1995-01-01

The current practice of calculating the seismic response is to perform the analysis of the primary structure (buildings) and the secondary systems (piping) separately. Earthquake input to the primary system in terms of a design response spectrum. An acceleration time history compatible with the design response spectrum is developed (a non-unique process) and primary system is analyzed to obtain the acceleration histories at the desired floors. Floor time histories are used for generating the corresponding instructure response spectrum (IRIS). The instructure response spectra are used as input at the supports of secondary systems. Further, in case of multiple supports, an envelope spectrum (introducing conservatism) is obtained from the individual support IRS. The effect of relative support motion is incorporated by a worst-case separate static analysis (adding to the conservatism). In the above method, mass interaction between the secondary and primary system is ignored, which may have significant effect at resonant frequencies (further adding to the conservatism). The calculated response may be an order of magnitude higher than they should be. Two computer programs, CREST and CREST-IRIS, were developed at Center for NUclear Power Plant Structures, Equipment and Piping. Any one of the two computer programs together with a piping analysis program can be used to perform an accurate coupled seismic analysis of piping systems. The two computer programs have been validated against the time history analysis for simple problems. In the present study, we have applied CREST to analyze two real-life piping systems. The piping analysis program used in this research is the commercial software PIPESTRESS, developed by DST Computer Services of Geneva, Switzerland. (author). 4 refs., 3 figs., 2 tabs

Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

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

International Nuclear Information System (INIS)

Slagis, G.C.

1996-01-01

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

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

Pipe support program at Pickering

International Nuclear Information System (INIS)

Sahazizian, L.A.; Jazic, Z.

1997-01-01

This paper describes the pipe support program at Pickering. The program addresses the highest priority in operating nuclear generating stations, safety. We present the need: safety, the process: managed and strategic, and the result: assurance of critical piping integrity. In the past, surveillance programs periodically inspected some systems, equipment, and individual components. This comprehensive program is based on a managed process that assesses risk to identify critical piping systems and supports and to develop a strategy for surveillance and maintenance. The strategy addresses all critical piping supports. Successful implementation of the program has provided assurance of critical piping and support integrity and has contributed to decreasing probability of pipe failure, reducing risk to worker and public safety, improving configuration management, and reducing probability of production losses. (author)

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 thermal fatigue evaluation methods of piping systems

International Nuclear Information System (INIS)

Kasahara, Naoto; Itoh, Takamoto; Okazaki, Masakazu; Okuda, Yukihiko; Kamaya, Masayuki; Nakamura, Akira; Nakamura, Hitoshi; Machida, Hideo; Matsumoto, Masaaki

2014-01-01

Nuclear piping has various kinds of thermal fatigue failure modes. Main causes of thermal loads are structural responses to fluid temperature changes during plant operation. These phenomena have complex mechanisms and many patterns, so that their problems still occur in spite of well-known issues. The guideline of the JSME (Japan Society of Mechanical Engineering) for estimation of thermal fatigue failures in piping system is employed as Japanese regulation. To improve this guideline, generation mechanisms of thermal load and fatigue failure have been investigated and summarized into the knowledgebase. And numerical simulation methods to replace experimental based methods were studied. Furthermore, probabilistic failure analysis approach with main influence parameters was investigated to be applied for the plant system safety. Thus, based on the knowledge, estimation methods revised from the JSME guideline were proposed. (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

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

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)

A simplified dynamic analysis for reactor piping systems under blowdown conditions

International Nuclear Information System (INIS)

Chen, M.M.

1975-01-01

In the design of pipelines in a nuclear power plant for blowdown conditions, is it customary to conduct dynamic analysis of the piping system to obtain the responses and the resulting stresses. Calculations are repeated for each design modification in piping geometry or supporting system until the design codes are met. The numerical calculations are, in general, very costly and time consuming. Until now, there have been no simple means for calculating the dynamic responses for the design. The proposed method reduces the dynamic calculation to a quasi-static one, and can be beneficially used for the preliminary design. The method is followed by a complete dynamical analysis to improve the final results. The new formulations greatly simplify the numerical computation and provide design guides. When used to design a given piping system, the method saved approximately one order of magnitude of computer time. The approach can also be used for other types of structures

Large eddy simulation on thermal fluid mixing in a T-junction piping system

Energy Technology Data Exchange (ETDEWEB)

Selvam, P. Karthick; Kulenovic, R.; Laurien, E. [Stuttgart Univ. (Germany). Inst fuer Kernenergie und Energiesysteme (IKE)

2014-11-15

High cycle thermal fatigue damage caused in piping systems is an important problem encountered in the context of nuclear safety and lifetime management of a Nuclear Power Plant (NPP). The T-junction piping system present in the Residual Heat Removal System (RHRS) is more vulnerable to thermal fatigue cracking. In this numerical study, thermal mixing of fluids at temperature difference (?T) of 117 K between the mixing fluids is analyzed. Large Eddy Simulation (LES) is performed with conjugate heat transfer between the fluid and structure. LES is performed based on the Fluid-Structure Interaction (FSI) test facility at University of Stuttgart. The results show an intense turbulent mixing of fluids downstream of T-junction. Amplitude of temperature fluctuations near the wall region and its corresponding frequency distribution is analyzed. LES is performed using commercial CFD software ANSYS CFX 14.0.

Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

International Nuclear Information System (INIS)

1975-10-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

CSNI specialist meeting on leak-before-break in nuclear reactor piping: proceedings

Energy Technology Data Exchange (ETDEWEB)

1984-08-01

On September 1 and 2, 1983, the CSNI subcommittee on primary system integrity held a special meeting in Monterey, California, on the subject of leak-before-break in nuclear reactor piping systems. The purpose of the meeting was to provide an international forum for the exchange of ideas, positions, and research results; to identify areas requiring additional research and development; and to determine the general attitude toward acceptance of the leak-before-break concept. The importance of the leak-before-break issue was evidenced by excellent attendance at the meeting and through active participation by the meeting attendees. Approximately 125 people representing fifteen different nations attended the meeting. The meeting was divided into four technical sessions addressing the following areas: Application of Piping Fracture Mechanics to Leak-Before Break, Leak Rate and Leak Detection, Leak-Before-Break Studies, Methods and Results, Current and Proposed Positions on Leak-Before-Break.

CSNI specialist meeting on leak-before-break in nuclear reactor piping: proceedings

International Nuclear Information System (INIS)

1984-08-01

On September 1 and 2, 1983, the CSNI subcommittee on primary system integrity held a special meeting in Monterey, California, on the subject of leak-before-break in nuclear reactor piping systems. The purpose of the meeting was to provide an international forum for the exchange of ideas, positions, and research results; to identify areas requiring additional research and development; and to determine the general attitude toward acceptance of the leak-before-break concept. The importance of the leak-before-break issue was evidenced by excellent attendance at the meeting and through active participation by the meeting attendees. Approximately 125 people representing fifteen different nations attended the meeting. The meeting was divided into four technical sessions addressing the following areas: Application of Piping Fracture Mechanics to Leak-Before Break, Leak Rate and Leak Detection, Leak-Before-Break Studies, Methods and Results, Current and Proposed Positions on Leak-Before-Break

Pipe inspection using the pipe crawler. Innovative technology summary report

International Nuclear Information System (INIS)

1999-05-01

The US Department of Energy (DOE) continually seeks safer and more cost-effective remediation technologies for use in the decontamination and decommissioning (D and D) of nuclear facilities. In several of the buildings at the Fernald Site, there is piping that was used to transport process materials. As the demolition of these buildings occur, disposal of this piping has become a costly issue. Currently, all process piping is cut into ten-foot or less sections, the ends of the piping are wrapped and taped to prevent the release of any potential contaminants into the air, and the piping is placed in roll off boxes for eventual repackaging and shipment to the Nevada Test Site (NTS) for disposal. Alternatives that allow for the onsite disposal of process piping are greatly desired due to the potential for dramatic savings in current offsite disposal costs. No means is currently employed to allow for the adequate inspection of the interior of piping, and consequently, process piping has been assumed to be internally contaminated and thus routinely disposed of at NTS. The BTX-II system incorporates a high-resolution micro color camera with lightheads, cabling, a monitor, and a video recorder. The complete probe is capable of inspecting pipes with an internal diameter (ID) as small as 1.4 inches. By using readily interchangeable lightheads, the same system is capable of inspecting piping up to 24 inches in ID. The original development of the BTX system was for inspection of boiler tubes and small diameter pipes for build-up, pitting, and corrosion. However, the system is well suited for inspecting the interior of most types of piping and other small, confined areas. The report describes the technology, its performance, uses, cost, regulatory and policy issues, and lessons learned

Pipe inspection using the pipe crawler. Innovative technology summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-05-01

The US Department of Energy (DOE) continually seeks safer and more cost-effective remediation technologies for use in the decontamination and decommissioning (D and D) of nuclear facilities. In several of the buildings at the Fernald Site, there is piping that was used to transport process materials. As the demolition of these buildings occur, disposal of this piping has become a costly issue. Currently, all process piping is cut into ten-foot or less sections, the ends of the piping are wrapped and taped to prevent the release of any potential contaminants into the air, and the piping is placed in roll off boxes for eventual repackaging and shipment to the Nevada Test Site (NTS) for disposal. Alternatives that allow for the onsite disposal of process piping are greatly desired due to the potential for dramatic savings in current offsite disposal costs. No means is currently employed to allow for the adequate inspection of the interior of piping, and consequently, process piping has been assumed to be internally contaminated and thus routinely disposed of at NTS. The BTX-II system incorporates a high-resolution micro color camera with lightheads, cabling, a monitor, and a video recorder. The complete probe is capable of inspecting pipes with an internal diameter (ID) as small as 1.4 inches. By using readily interchangeable lightheads, the same system is capable of inspecting piping up to 24 inches in ID. The original development of the BTX system was for inspection of boiler tubes and small diameter pipes for build-up, pitting, and corrosion. However, the system is well suited for inspecting the interior of most types of piping and other small, confined areas. The report describes the technology, its performance, uses, cost, regulatory and policy issues, and lessons learned.

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...

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)

Computer simulation of LMFBR piping systems

International Nuclear Information System (INIS)

A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

1977-01-01

Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

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.

Radiation transmission pipe thickness measurement system

International Nuclear Information System (INIS)

Higashi, Yasuhiko

2010-01-01

Fuji Electric Systems can be measured from the outer insulation of the transmission Characteristics and radiation detection equipment had been developed that can measure pipe wall thinning in plant and running, the recruitment of another three-beam calculation method by pipe thickness measurement system was developed to measure the thickness of the pipe side. This equipment has been possible to measure the thickness of the circumferential profile of the pipe attachment by adopting automatic rotation. (author)

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

International Nuclear Information System (INIS)

Jiang Zhenbiao; Li Guanying; Liu Zhicheng

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

Nuclear power system

International Nuclear Information System (INIS)

Yampolsky, J.S.; Cavallaro, L.; Paulovich, K.F.; Schleicher, R.W.

1989-01-01

This patent describes an inherently safe modular nuclear power system for producing electrical power at acceptable efficiency levels using working fluids at relatively low temperatures and pressures. The system comprising: a reactor module for heating a first fluid; a heat exchanger module for transferring heat from the first fluid to a second fluid; a first piping system effecting flow of the first fluid in a first fluid circuit successively through the reactor module and the heat exchanger module; a power conversion module comprising a turbogenerator driven by the second fluid, and means for cooling the second fluid upon emergence thereof from the turbogenerator; a second piping system comprising means for effecting flow of the second fluid in a second fluid circuit successively through the heat exchanger module and the power conversion module; and a plurality of pits for receiving the modules

Pipe Crawler internal piping characterization system. Deactivation and decommissioning focus area. Innovative Technology Summary Report

International Nuclear Information System (INIS)

1998-02-01

Pipe Crawler reg-sign is a pipe surveying system for performing radiological characterization and/or free release surveys of piping systems. The technology employs a family of manually advanced, wheeled platforms, or crawlers, fitted with one or more arrays of thin Geiger Mueller (GM) detectors operated from an external power supply and data processing unit. Survey readings are taken in a step-wise fashion. A video camera and tape recording system are used for video surveys of pipe interiors prior to and during radiological surveys. Pipe Crawler reg-sign has potential advantages over the baseline and other technologies in areas of cost, durability, waste minimization, and intrusiveness. Advantages include potentially reduced cost, potential reuse of the pipe system, reduced waste volume, and the ability to manage pipes in place with minimal disturbance to facility operations. Advantages over competing technologies include potentially reduced costs and the ability to perform beta-gamma surveys that are capable of passing regulatory scrutiny for free release of piping systems

Piping engineering and operation

International Nuclear Information System (INIS)

1993-01-01

The conference 'Piping Engineering and Operation' was organized by the Institution of Mechanical Engineers in November/December 1993 to follow on from similar successful events of 1985 and 1989, which were attended by representatives from all sectors of the piping industry. Development of engineering and operation of piping systems in all aspects, including non-metallic materials, are highlighted. The range of issues covered represents a balance between current practices and implementation of future international standards. Twenty papers are printed. Two, which are concerned with pressurized pipes or steam lines in the nuclear industry, are indexed separately. (Author)

Elastic-plastic dynamic behavior of guard pipes due to sudden opening of longitudinal cracks in the inner pipe and crash to the guard pipe wall

International Nuclear Information System (INIS)

Theuer, E.; Heller, M.

1979-01-01

Integrity of guard pipes is an important parameter in the design of nuclear steam supply systems. A guard pipe shall withstand all kinds of postulated inner pipe breaks without failure. Sudden opening of a crack in the inner pipe and crash of crack borders to the guard pipe wall represent a shock problem where complex phenomena of dynamic plastification as well as dynamic behavior of the entire system have to be taken in consideration. The problem was analyzed by means of Finite Element computation using the general purpose program MARC. Equation of motion was resolved by direct integration using the Newmark β-operator. Analysis shows that after 1,2 m sec crack borders touch the guard pipe wall for the first time. At this moment a considerable amount of local plastification appears in the inner pipe wall, while the guard pipe is nearly unstressed. After initial touching, the crack borders begin to slip along the guard pipe wall. Subsequently, a short withdrawal of the crack borders and a new crash occur, while the inner pipe rolls along the guard pipe wall. The analysis procedure described is suitable for designing numerous guard pipe geometries as well as U-Bolt restraint systems which have to withstand high-energy pipe rupture impact. (orig.)

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.

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)

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

The IPIRG-1 pipe system fracture tests: Experimental results

International Nuclear Information System (INIS)

Scott, P.; Olson, R.J.; Wilkowski, G.M.

1994-01-01

As part of the First International Piping Integrity Research Group (IPIRG-1) program, six dynamic pipe system experiments were conducted. The objective of these experiments was to generate experimental data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system subjected to combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The pipe system evaluated was an expansion loop with over 30 m (100 feet) of 16-inch nominal diameter Schedule 100 pipe. The experimental facility was equipped with special hardware to ensure that system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe system experiments. The uncracked-pipe experiment was conducted to evaluate the piping system damping and natural frequency characteristics. The cracked-pipe experiments were conducted to evaluate the fracture behavior, piping system response, and fracture stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided the tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Key results from the six pipe system experiments and material characterization efforts are presented. Detailed analyses will be published in a companion paper

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)

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

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.)

Comparison of multiple support excitation solution techniques for piping systems

International Nuclear Information System (INIS)

Sterkel, H.P.; Leimbach, K.R.

1980-01-01

Design and analysis of nuclear power plant piping systems exposed to a variety of dynamic loads often require multiple support excitation analysis by modal or direct time integration methods. Both methods have recently been implemented in the computer program KWUROHR for static and dynamic analysis of piping systems, following the previous implementation of the multiple support excitation response spectrum method (see papers K 6/15 and K 6/15a of the SMiRT-4 Conference). The results of multiple support excitation response spectrum analyses can be examined by carrying out the equivalent time history analyses which do not distort the time phase relationship between the excitations at different support points. A frequent point of discussion is multiple versus single support excitation. A single support excitation analysis is computationally straightforward and tends to be on the conservative side, as the numerical results show. A multiple support excitation analysis, however, does not incur much more additional computer cost than the expenditure for an initial static solution involving three times the number, L, of excitation levels, i.e. 3L static load cases. The results are more realistic than those from a single support excitation analysis. A number of typical nuclear plant piping systems have been analyzed using single and multiple support excitation algorithms for: (1) the response spectrum method, (2) the modal time history method via the Wilson, Newmark and Goldberg integration operators and (3) the direct time history method via the Wilson integration operator. Characteristic results are presented to compare the computational quality of all three methods. (orig.)

Internal testing of pipe systems with IRIS inspection system

International Nuclear Information System (INIS)

1986-01-01

The internal piping inspection system IRIS allows inside testing of pipes with an internal diameter of NW 70 as a minimum, and of any horizontal or vertical layout of the piping system. Visual testing is done by means of an integrated CCD video system with high resolution power. Technical data are given and examples of applications, in the German and English language. (DG) [de

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.)

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 carbon steel with superior resistance to wall thinning and fracture for nuclear piping system

International Nuclear Information System (INIS)

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

2010-07-01

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

Design Evaluation of a Piping System in the SELFA Sodium Test Facility

Energy Technology Data Exchange (ETDEWEB)

Son, Seok-Kwon; Jo, Young-Chul; Lee, Hyeong-Yeon; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this study, design evaluations on the SELFA piping system has been conducted according to the ASME B31.1 and RCC-MRx RD-3600. The conservatism of the two codes was quantified based on the evaluation results. It was shown that B31.1 was more conservative for the sustained loads while less conservative for thermal expansion loads when compare with those of RD-3600. However, all the evaluation results according to the two codes were within the code allowables. There are two main piping systems in the SELFA test loop. In this study, the integrity of the SELFA piping system has been evaluated according to the two design-by-rule (DBR) codes of ASME B31.1 and RCC-MRx RD-3600. B31.1 is an industry design code for power piping while RD-3600 is a class 3 nuclear DBR code. The conservatism of the two codes was quantified based on the evaluation results as per the two DBR codes. The sodium test facility of the SELFA is under construction at KAERI for the investigation of thermo-hydraulic behavior of finned-tube sodium-to-air heat exchanger.

Development of total systems of piping stress analysis and evaluation: ISAPPS

International Nuclear Information System (INIS)

Oki, Teizaburo; Koyanagi, Ryoichi; Fukuda, Masanao

1978-01-01

IHI has developed the systems of piping stress analysis and evaluation: ISAPPS (IHI Stress Analysis Program for Piping Systems), which are further described in this paper. In addition, the results of structural analysis and heat transfer analysis were confirmed. An example of stress evaluation in accordance with the modified ASME Code Sec. III is shown. ISAPPS consists of the following seven parts, and is designed for easy adoption of other programs by making modifications. 1. Piping design oriented language programs 2. Structural analysis programs 3. Isometric plotting programs 4. Multi-file dumping program 5. Load combination program 6. Heat transfer program 7. Stress evaluation programs As one of the examples of structural analysis programs, IHI make use of the modified SAP IV developed by the University of California. Evaluations of stresses are performed in accordance with: 1. ASME Boiler and Pressure Vessel Code, Sec. III Class 1, 2 and 3 2. ANSI Code, B31.1 and B31.3 3. MITI (Ministry of International Trade and Industry ) Code ISAPPS is very useful for design of nuclear and chemical pipings and so on. (author)

Heat pipes and heat pipe exchangers for heat recovery systems

Energy Technology Data Exchange (ETDEWEB)

Vasiliev, L L; Grakovich, L P; Kiselev, V G; Kurustalev, D K; Matveev, Yu

1984-01-01

Heat pipes and heat pipe exchangers are of great importance in power engineering as a means of recovering waste heat of industrial enterprises, solar energy, geothermal waters and deep soil. Heat pipes are highly effective heat transfer units for transferring thermal energy over large distance (tens of meters) with low temperature drops. Their heat transfer characteristics and reliable working for more than 10-15 yr permit the design of new systems with higher heat engineering parameters.

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.

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

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

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.

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

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

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)

An experimental study of an energy absorbing restrainer for piping systems

International Nuclear Information System (INIS)

Sone, A.; Suzuki, K.

1989-01-01

Recently, in the seismic design methodology of the piping systems in nuclear power plants, new and improved design criteria and calculation techniques which will lead to more reliable and cost saving design products have been investigated. For instance, problems for reducing the snubbers in nuclear power plants which provide high costs for their inspections and maintenances and related flexible design problems for the dynamic piping systems have been investigated. Thus, in order to replace snubbers, various types of alternative supporting devices such as dynamic absorbers, gapped support and energy absorbing support devices have been proposed. A number of energy absorbing restrainers have been designed in Japan and United-States by allowing yield to occur during strong earthquakes. Advantages and disadvantages of these restrainers were examined analytically and experimentally. In order to overcome the disadvantages, the authors introduced new absorbing material LSPZ (laminated super plastic zinc) in which SPZ is expected to have reliable ductility and also efficient energy absorbability still under the normal temperature condition. This paper is devoted to an experimental works for this updated absorbing restrainer

An investigation of elastic-plastic seismic analysis of piping systems under high level of earthquake motion

International Nuclear Information System (INIS)

Liu, T.H.; Patel, R.B.; Condrac, R.

1993-01-01

The current design by rules of the ASME Section III Code for the nuclear power plant piping system is principally based on the elastic design concept Such design often results in a more rigid piping system, structurally, that may not be so desirable from the viewpoint of long term plant operation. The so called 'elastic design' approach has failed to utilize the ductility that steel pipe exhibits, and therefore, the resulting system maintains a great deal of reserve margin in seismic design. This study does not attempt to assess the amount of this reserve margin but provides some findings and discussions with respect to dynamic inelastic analysis results in the piping system design. Using a test correlation analysis it was found that, while the analytical tools that exist are conservative for low strain levels, further studies with loadings at high strain levels are recommended for a more reasonable design. (author)

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)

Evaluation of clamp effects on LMFBR piping systems

International Nuclear Information System (INIS)

Jones, G.L.

1980-01-01

Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness

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

Czech Academy of Sciences Publication Activity Database

Medek, Jiří; Weishauptová, Zuzana

2009-01-01

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

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.

Practical application of equivalent linearization approaches to nonlinear piping systems

International Nuclear Information System (INIS)

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

1995-01-01

The use of mechanical energy absorbers as an alternative to conventional hydraulic and mechanical snubbers for piping supports has attracted a wide interest among researchers and practitioners in the nuclear industry. The basic design concept of energy absorbers (EA) is to dissipate the vibration energy of piping systems through nonlinear hysteretic actions of EA exclamation point s under design seismic loads. Therefore, some type of nonlinear analysis needs to be performed in the seismic design of piping systems with EA supports. The equivalent linearization approach (ELA) can be a practical analysis tool for this purpose, particularly when the response approach (RSA) is also incorporated in the analysis formulations. In this paper, the following ELA/RSA methods are presented and compared to each other regarding their practice and numerical accuracy: Response approach using the square root of sum of squares (SRSS) approximation (denoted RS in this paper). Classical ELA based on modal combinations and linear random vibration theory (denoted CELA in this paper). Stochastic ELA based on direct solution of response covariance matrix (denoted SELA in this paper). New algorithms to convert response spectra to the equivalent power spectral density (PSD) functions are presented for both the above CELA and SELA methods. The numerical accuracy of the three EL are studied through a parametric error analysis. Finally, the practicality of the presented analysis is demonstrated in two application examples for piping systems with EA supports

Subprogram Calculating The Distance Between Pipe And Plane For Automatic Piping System Design

International Nuclear Information System (INIS)

Satmoko, Ari

2001-01-01

DISTLNPL subprogram was created using Auto LISP software. This subprogram is planned to complete CAPD (Computer Aided Piping Design) software being developed. The CAPD works under the following method: suggesting piping system line and evaluating whether any obstacle allows the proposed line to be constructed. DISTLNPL is able to compute the distance between pipe and any equipment having plane dimension such as wall, platform, floors, and so on. The pipe is modeled by using a line representing its axis, and the equipment is modeled using a plane limited by some lines. The obtained distance between line and plane gives information whether the pipe crosses the equipment. In the case of crashing, the subprogram will suggest an alternative point to be passed by piping system. So far, DISTLNPL has not been able to be accessed by CAPD yet. However, this subprogram promises good prospect in modeling wall, platform, and floors

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

International Nuclear Information System (INIS)

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

2005-01-01

The optimal location of supports for the reduction of vibration in piping systems is an interesting structural problem, that can be approached also with the methods used in the updating parameters problems, especially when the cause of the vibration is due to a resonance or a ineffective damping level, and the shift of a critical frequency is an effective remedy. In the proposed paper a frequency domain method of Finite Element model updating is used in order to properly locate the natural frequencies of a given piping system, starting from the nominal condition. The method considers modal parameters, and enables to directly update the physical quantities of the finite element model, i.e., in the considered application, the stiffness and damping of the supporting devices. The model of the structure is made up by beam finite elements, lumped springs and masses, and rigid links. The general procedure is applied here updating the stiffness and location of the supports of the piping according to the comparison among the current frequencies of the piping systems structure and the ones that are required, according to specific requirements. Application to a system similar to a portion of an actual AP1000 (a new Advanced Passive Nuclear Reactor) piping-support layout is presented. (authors)

Inspection indications, stress corrosion cracks and repair of process piping in nuclear materials production reactors

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; West, S.L.; Nelson, D.Z.

1991-01-01

Ultrasonic inspection of Schedule 40 Type 304 stainless steel piping in the process water system of the Savannah River Site reactors has provided indications of discontinuities in less than 10% of the weld heat affected zones. Pipe sections containing significant indications are replaced with Type 304L components. Post removal metallurgical evaluation showed that the indications resulted from stress corrosion cracking in weld heat-affected zones and that the overall weld quality was excellent. The evaluation also revealed weld fusion zone discontinuities such as incomplete penetration, incomplete fusion, inclusions, underfill at weld roots and hot cracks. Service induced extension of these discontinuities was generally not significant although stress corrosion cracking in one weld fusion zone was noted. One set of UT indications was caused by metallurgical discontinuities at the fusion boundary of an extra weld. This extra weld, not apparent on the outer pipe surface, was slightly overlapping and approximately parallel to the weld being inspected. This extra weld was made during a pipe repair, probably associated with initial construction processes. The two nearly parallel welds made accurate assessment of the UT signal difficult. The implications of these observations to the inspection and repair of process water systems of nuclear reactors is discussed

Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms

International Nuclear Information System (INIS)

Datta, Debashis

2010-02-01

A nuclear piping system is composed of several straight pipes and elbows joined by welding. These weld sections are usually the most susceptible failure parts susceptible to various degradation mechanisms. Whereas a specific location of a reactor piping system might fail by a combination of different aging mechanisms, e.g. fatigue and/or stress corrosion cracking, the majority of the piping probabilistic fracture mechanics (PFM) codes can only consider a single aging mechanism at a time. So, a probabilistic fracture mechanics computer code capable of considering multiple aging mechanisms was developed for an accurate failure analysis of each specific component of a nuclear piping section. The newly proposed crack morphology based probabilistic leak flow rate module is introduced in this code to separately treat fatigue and SCC type cracks. Improved models e.g. stressors models, elbow failure model, SIFs model, local seismic occurrence probability model, performance based crack detection models, etc., are also included in this code. Recent probabilistic fatigue (S-N) and SCC crack initiation (S-T) and subsequent crack growth rate models are coded. An integrated probabilistic risk assessment and probabilistic fracture mechanics methodology is proposed. A complete flow chart regarding the combined aging mechanism model is presented. The combined aging mechanism based module can significantly reduce simulation efforts and time. Two NUREG benchmark problems, e.g. reactor pressure vessel outlet nozzle section and a surge line elbow located just below the pressurizer are reinvestigated by this code. The results showed that, contribution of pre-existing cracks in addition to initiating cracks, can significantly increase the overall failure probability. Inconel weld location of reactor pressure vessel outlet nozzle section showed the weakest point in terms of relative through-wall leak failure probability in the order of about 10 -2 at the 40-year plant life. Considering

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.)

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

System and Method for Traversing Pipes

Science.gov (United States)

Graf, Jodi (Inventor); Pettinger, Ross (Inventor); Azimi, Shaun (Inventor); Magruder, Darby (Inventor); Ridley, Justin (Inventor); Lapp, Anthony (Inventor)

2017-01-01

A system and method is provided for traversing inside one or more pipes. In an embodiment, a fluid is injected into the one or more pipes thereby promoting a fluid flow. An inspection device is deployed into the one or more pipes at least partially filled with a flowing fluid. The inspection device comprises a housing wherein the housing is designed to exploit the hydrokinetic effects associated with a fluid flow in one or more pipes as well as maneuver past a variety of pipe configurations. The inspection device may contain one or more sensors capable of performing a variety of inspection tasks.

Determination of limits for smallest detectable and largest subcritical leakage cracks in piping systems

International Nuclear Information System (INIS)

Bieselt, R.; Wolf, M.

1995-01-01

Nuclear power plant piping systems - those still in their original as-built condition as well as upgraded designs - are subject to safety analysis. In order to limit the consequences of postulated piping failures, the basic safety concept incorporating rupture preclusion criteria is applied to specific high-energy piping systems. Leak-before-break analyses are also conducted within the framework of this concept. These analyses serve to determine the potential consequences of jet and reaction forces due to maximum subcritical leakage cracks while also establishing the minimum crack sizes that would be reliably detectable by the leakage rates resulting from these cracks. The boundary conditions for these analyses are not clearly defined. Using various examples as a basis, this paper presents and discusses how the leak-before-break concept can be applied. (orig.)

Seismic evaluation of piping systems using screening criteria

International Nuclear Information System (INIS)

Campbell, R.D.; Landers, D.F.; Minichiello, J.C.; Slagis, G.C.; Antaki, G.A.

1994-01-01

This document may be used by a qualified review team to identify potential sources of seismically induced failure in a piping system. Failure refers to the inability of a piping system to perform its expected function following an earthquake, as defined in Table 1. The screens may be used alone or with the Seismic Qualification Utility Group -- Generic Implementation Procedure (SQUG-GIP), depending on the piping system's required function, listed in Table 1. Features of a piping system which do not the screening criteria are called outliers. Outliers must either be resolved through further evaluations, or be considered a potential source of seismically induced failure. Outlier evaluations, which do not necessarily require the qualification of a complete piping system by stress analysis, may be based on one or more of the following: simple calculations of pipe spans, search of the test or experience data, vendor data, industry practice, etc

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

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 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)

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.)

Comparison and evaluation of flexible and stiff piping systems

International Nuclear Information System (INIS)

Hahn, W.; Tang, H.T.; Tang, Y.K.

1983-01-01

An experimental and numerical study was performed on a piping system, with various support configurations, to assess the difference in piping response for flexible and stiff piping systems. Questions have arisen concerning a basic design philosophy employed in present day piping designs. One basic question is, the reliability of a flexible piping system greater than that of a stiff piping system by virtue of the fact that a flexible system has fewer snubber supports. With fewer snubbers, the pipe is less susceptible to inadvertent thermal stresses introduced by snubber malfunction during normal operation. In addition to the technical issue, the matter of cost savings in flexible piping system design is a significant one. The costs associated with construction, in-service inspection and maintenance are all significantly reduced by reducing the number of snubber supports. The evaluation study, sponsored by the Electric Power Research Institute, was performed on a boiler feedwater line at Consolidated Edison's Indian Point Unit 1. In this study, the boiler feedwater line was tested and analyzed with two fundamentally different support systems. The first system was very flexible, employing rod and spring hangers, and represented the 'old' design philosophy. The pipe system was very flexible with this support system, due to the long pipe span lengths between supports and the fact that there was only one lateral support. This support did not provide much restraint since it was near an anchor. The second system employed strut and snubber supports and represented the 'modern' design philosophy. The pipe system was relatively stiff with this support system, primarily due to the increased number of supports, including lateral supports, thereby reducing the pipe span lengths between supports. The second support system was designed with removable supports to facilitate interchange of the supports with different support types (i.e., struts, mechanical snubbers and hydraulic

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.

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

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

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.)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-03-15

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

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

Water Hammer Analysis using RELAP5/MOD 3.3 for Yonggwang Nuclear Power Unit 1 and 2 Blowdown System

International Nuclear Information System (INIS)

Lee, Sang Il; Kim, Hea Zoo; Chu, Jung Ho; Ahn, Se Hong; Jung, Chang Ho

2010-01-01

Water hammer can be defined as a rapid pressure step occurring in the liquid in a closed pipe caused by a sudden change in the liquid velocity. This pressure acts for a period which is twice the transit time of sonic wave in the pipe. Generally, water hammer can occur in any thermal-hydraulic systems like nuclear power plant and is extremely dangerous for nuclear power plant piping system since, if the pressure induced exceeds the pressure range of the pipe given by the manufacturer, it can lead to the failure of the piping system integrity. For Yonggwang nuclear power unit 1 and 2, water hammer occurred repeatedly on the outlet piping of regenerative heat exchanger of steam generator blowdown system. Thus, design modification was performed to prevent the water hammer and the analysis of effect on water hammer before and after design modification was performed to verify the validity of the design modification

Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

International Nuclear Information System (INIS)

Lowry, W.

1994-01-01

The objective for the development of the Pipe Explorer trademark radiological characterization system is to achieve a cost effective, low risk means of characterizing gamma radioactivity on the inside surface of pipes. The unique feature of this inspection system is the use of a pneumatically inflated impermeable membrane which transports the detector into the pipe as it inverts. The membrane's internal air pressure tows the detector and tether through the pipe. This mechanism isolates the detector and its cabling from the contaminated surface, yet allows measurement of radioactive emissions which can readily penetrate the thin plastic membrane material (such as gamma and high energy beta emissions). In Phase 1, an initial survey of DOE facilities was conducted to determine the physical and radiological characteristics of piping systems. The inverting membrane deployment system was designed and extensively tested in the laboratory. A range of membrane materials was tested to evaluate their ruggedness and deployment characteristics. Two different sizes of gamma scintillation detectors were procured and tested with calibrated sources. Radiation transport modeling evaluated the measurement system's sensitivity to detector position relative to the contaminated surface, the distribution of the contamination, background gamma levels, and gamma source energy levels. In the culmination of Phase 1, a field demonstration was conducted at the Idaho National Engineering Laboratory's Idaho Chemical Processing Plant. The project is currently in transition from Phase 1 to Phase 2, where more extensive demonstrations will occur at several sites. Results to date are discussed

Experimental benchmark for piping system dynamic-response analyses

International Nuclear Information System (INIS)

1981-01-01

This paper describes the scope and status of a piping system dynamics test program. A 0.20 m(8 in.) nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Particular attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed

Experimental benchmark for piping system dynamic response analyses

International Nuclear Information System (INIS)

Schott, G.A.; Mallett, R.H.

1981-01-01

The scope and status of a piping system dynamics test program are described. A 0.20-m nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed. 3 refs

Degradation mechanisms of small scale piping systems

International Nuclear Information System (INIS)

Bartonicek, J.; Koenig, G.; Blind, D.

1996-01-01

Operational experience shows that many degradation mechanisms can have an effect on small-scale piping systems. We can see from the analyses carried out that the degradation which has occurred is primarily linked with the fact that these piping systems were classified as being of low safety relevance. This is mainly due to such components being classified into low safety relevance category at the design stage, as well as to the low level of operational monitoring. Since in spite of the variety of designs and operational modes the degradation mechanisms detected may be attributed to the piping systems, we can make decisive statements on how to avoid such degradation mechanisms. Even small-scale piping systems may achieve guaranteed integrity in such cases by taking the appropriate action. (orig.) [de

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.

A critical review on the application of elastic-plastic fracture mechanics to nuclear pressure vessel and piping systems

International Nuclear Information System (INIS)

Scarth, D.A.; Kim, Y.J.; Vanderglas, M.L.

1985-10-01

A comprehensive literature survey on the application of Elastic-Plastic Fracture Mechanics to the assessment of the structural integrity of nuclear pressure vessels and piping is presented. In particular, the J-integral/Tearing Modulus (J/T) approach and the Failure Assessment Diagram (FAD) are covered in detail because of their general suitability for use in Ontario Hydro. (25 refs.)

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

Tensile and fracture properties of primary heat transport system piping material

International Nuclear Information System (INIS)

Singh, P.K.; Chattopadhyay, J.; Kushwaha, H.S.

1997-07-01

The fracture mechanics calculations in leak-before-break analysis of nuclear piping system require material tensile data and fracture resistance properties in the form of J-R curve. There are large variations in fracture parameters due to variation in chemical composition and process used in making the steel components. Keeping this in view, a comprehensive program has been planned to generate the material data base for primary heat transport system piping using the specimens machined from actual pipes used in service. The material under study are SA333 Gr.6 (base as well as weld) and SA350 LF2 (base). Since the operating temperatures of 500 MWe Indian PHWR PHT system piping range from 260 degC to 304 degC the test temperature chosen are 28 degC, 200 degC, 250 degC and 300 degC. Tensile and compact tension specimens have been fabricated from actual pipe according to ASTM standard. Fracture toughness of base metal has been observed to be higher compared to weld metal in SA333 Gr.6 material for the temperature under consideration. Fracture toughness has been observed to be higher for LC orientation (notch in circumferential direction) compared to CL orientation (notch is in longitudinal direction) for the temperature range under study. Fracture toughness value decreases with increase in temperature for the materials under study. Finally, chemical analysis has been carried out to investigate the reason for high toughness of the material. It has been concluded that low percentage of carbon and nitrogen, low inclusion rating and fine grain size has enhanced the fracture toughness value

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

International Nuclear Information System (INIS)

Stoykovich, M.

1982-01-01

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

Pipe Explorer surveying system. Innovative technology summary report

International Nuclear Information System (INIS)

1999-06-01

The US Department of Energy's (DOE) Chicago Operations Office and the DOE's Federal Energy Technology Center (FETC) developed a Large Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial decontamination and decommissioning (D and D) technologies in comparison with current baseline technologies. The Pipe Explorer trademark system was developed by Science and Engineering Associates, Inc. (SEA), Albuquerque, NM as a deployment method for transporting a variety of survey tools into pipes and ducts. Tools available for use with the system include alpha, beta and gamma radiation detectors; video cameras; and pipe locator beacons. Different versions of this technology have been demonstrated at three other sites; results of these demonstrations are provided in an earlier Innovative Technology Summary Report. As part of a D and D project, characterization radiological contamination inside piping systems is necessary before pipes can be recycled, remediated or disposed. This is usually done manually by surveying over the outside of the piping only, with limited effectiveness and risk of worker exposure. The pipe must be accessible to workers, and embedded pipes in concrete or in the ground would have to be excavated at high cost and risk of exposure to workers. The advantage of the Pipe Explorer is its ability to perform in-situ characterization of pipe internals

Small pipe characterization system (SPCS) conceptual design

Energy Technology Data Exchange (ETDEWEB)

Anderson, M.O.; Ferrante, T.A.; McKay, M.D.

1995-01-01

Throughout the Department of Energy (DOE) complex there are many facilities that have been identified for Decontamination and Decommissioning (D&D). As processes are terminated or brought off-line, facilities are placed on the inactive list, and facility managers and site contractors are required to assure a safe and reliable decommissioning and transition of these facilities to a clean final state. Decommissioning of facilities requires extensive reliable characterization, decontamination and in some cases dismantlement. Characterization of piping systems throughout the DOE complex is becoming more and more necessary. In addition to decommissioning activities, characterization activities are performed as part of surveillance and maintenance (S&M). Because of the extent of contamination, all inactive facilities require some type of S&M. These S&M activities include visual assessment, equipment and material accounting, and maintenance. The majority of the inactive facilities have piping systems 3 inches or smaller that are inaccessible because they are contaminated, imbedded in concrete, or run through hot cells. Many of these piping systems have been inactive for a number of years and there exists no current system condition information or the historical records are poor and/or missing altogether. Many of these piping systems are placed on the contaminated list, not because of known contamination, but because of the risk of internal contamination. Many of the piping systems placed on the contamination list may not have internal contamination. Because there is a potential however, they are treated as such. The cost of D&D can be greatly reduced by identifying and removing hot spot contamination, leaving clean piping to be removed using conventional methods. Accurate characterization of these piping systems is essential before, during and after all D&D activities.

Small pipe characterization system (SPCS) conceptual design

International Nuclear Information System (INIS)

Anderson, M.O.; Ferrante, T.A.; McKay, M.D.

1995-01-01

Throughout the Department of Energy (DOE) complex there are many facilities that have been identified for Decontamination and Decommissioning (D ampersand D). As processes are terminated or brought off-line, facilities are placed on the inactive list, and facility managers and site contractors are required to assure a safe and reliable decommissioning and transition of these facilities to a clean final state. Decommissioning of facilities requires extensive reliable characterization, decontamination and in some cases dismantlement. Characterization of piping systems throughout the DOE complex is becoming more and more necessary. In addition to decommissioning activities, characterization activities are performed as part of surveillance and maintenance (S ampersand M). Because of the extent of contamination, all inactive facilities require some type of S ampersand M. These S ampersand M activities include visual assessment, equipment and material accounting, and maintenance. The majority of the inactive facilities have piping systems 3 inches or smaller that are inaccessible because they are contaminated, imbedded in concrete, or run through hot cells. Many of these piping systems have been inactive for a number of years and there exists no current system condition information or the historical records are poor and/or missing altogether. Many of these piping systems are placed on the contaminated list, not because of known contamination, but because of the risk of internal contamination. Many of the piping systems placed on the contamination list may not have internal contamination. Because there is a potential however, they are treated as such. The cost of D ampersand D can be greatly reduced by identifying and removing hot spot contamination, leaving clean piping to be removed using conventional methods. Accurate characterization of these piping systems is essential before, during and after all D ampersand D activities

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.)

A study on probabilistic fracture mechanics for nuclear pressure vessels and piping

International Nuclear Information System (INIS)

Yagawa, Genki; Yoshimura, Shinobu

1997-01-01

This paper describes some recent research activities on probabilistic fracture mechanics (PFM) for nuclear pressure vessels and piping (PV and P) performed by the RC111 research committee of the Japan Society of Mechanical Engineers (JSME) under a subcontract of the Japan Atomic Energy Research Institute (JAERI). To establish standard procedures for evaluating failure probabilities of nuclear PV and P, we have set up the following three kinds of PFM round-robin problems on: (a) primary piping under normal operating conditions, (b) aged reactor pressure vessel (RPV) under normal and upset operating conditions, and (c) aged RPV under pressurised thermal shock (PTS) events. The basic problems of the last one are chosen from some US benchmark problems such as EPRI (Electric Power Research Institute) and US NRC (Nuclear Regulatory Commission) joint PTS benchmark problems. This paper summarizes some sensitivity studies on the three kinds of problems mainly varying material properties such as flow stress, fracture toughness, fatigue crack growth rate, Cu content. Employed in this study are the PFM computer codes developed in Japan and USA. Failure probabilities of nuclear PV and P are quantitatively discussed in detail. (author)

Pipe damping: experimental results from laboratory tests in the seismic frequency range

International Nuclear Information System (INIS)

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

1986-06-01

The Idaho National Engineering Laboratory (INEL) has been conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for the seismic analysis of nuclear piping systems. As part of this program, a 5-in. piping system was tested by the INEL, and data from USNRC/EPRI piping vibration tests at the ANCO Engineers facility were evaluated. These systems were subjected to various types of excitation methods and magnitudes, the support configurations were varied, and the effects of pipe insulation and internal pressure were investigated on the INEL system. The INEL has used several different methods to reduce the data to determine the damping in both these piping systems under the various test conditions. It was concluded that at representative seismic excitation levels, pressure was not a contributing factor, but the supports, insulation, and magnitude of response all were major influences contributing to damping. These tests are part of the ongoing program to determine how various parameters and data reduction methods affect piping system damping. The evaluation of all relevant test results, including these two series, will potentially lead to revised damping guidelines for the seismic analysis of nuclear plants, making them safer, less costly, and easier to inspect and maintain. The test results as well as accompanying evaluations and recommendations are presented in this report. 27 refs., 72 figs., 13 tabs

Inspection of secondary cooling system piping of JMTR

International Nuclear Information System (INIS)

Hanawa, Yoshio; Izumo, Hironobu; Fukasaku, Akitomi; Nagao, Yoshiharu; Kawamura, Hiroshi

2008-06-01

Piping condition was inspected form the view point of long term utilization before the renewal work of the secondary cooling system in the JMTR on FY 2008. As the result, it was confirmed that cracks, swellings and exfoliations in inner lining of the piping could be observed, and corrosion, which was reached by piping ingot, or decrease of piping thickness could hardly be observed. It was therefore confirmed that the strength or the functionality of the piping had been maintained by usual operation and maintenance. Repair of inner lining of the piping during the refurbishment of the JMTR is necessary to long term utilization of the secondary cooling system after restart of the JMTR from the view point of preventive maintenance. In addition, a periodic inspection of inner lining condition is necessary after repair of the piping. (author)

A regulatory perspective on appropriate seismic loading stress criteria for advanced light water reactor piping systems

International Nuclear Information System (INIS)

Terao, D.

1995-01-01

In the foregoing sections, the author has discussed the NRC staff's perspective on the evolving seismic design criteria for piping systems. He also addressed the need for developing seismic loading stress criteria and provided several recommendations and considerations for ensuring piping functional capability, pressure integrity, and structural integrity. Overall, the general consensus in the NRC staff is that in the past several years, many initiatives have been developed and implemented by the industry and the NRC staff to reduce the excessive conservatisms that might have existed in nuclear piping system design criteria. The regulations, regulatory guides, and Standard Review Plan have been (or are currently in the process of being) revised to reflect these initiatives in an effort to produce requirements and guidelines that will continue to result in a safe and practical design of piping systems. However, further proposals to reduce margins are continually being submitted to the ASME Boiler and Pressure Vessel Code and the NRC for review and approval. Improvements to the piping seismic design criteria are always encouraged, but there is a point at which the benefits might be outweighed by drawbacks. Because of this rapidly evolving situation the need exists for the industry and the NRC staff to develop a course of action to ensure that piping seismic design criteria for future ALWR plants will result in piping system designs that provide adequate safety margins and practical designs at a reasonable cost

Evaluation on the thermal-hydraulic behavior of condensation pool and piping system

Energy Technology Data Exchange (ETDEWEB)

Kwon, Soon Bum; Lee, B. E.; Baek, S. C.; Joo, S. Y.; Lee, D. E.; Woo, S. W. [Kyungpook National Univ., Daegu (Korea, Republic of)

2002-03-15

The In-containment Refueling Water Storage Tank (IRWST) has the function of heat sink, when steam is released from the pressurizer. The hydrodynamic behaviors occurring at the piping system and sparger are very complex because of the wide variety of operating conditions and the complex geometry. Hydrodynamic behavior when air is discharged through a sparger in a condensation pool is investigated using CFD techniques in the present study. The effect of pressure acting on the sparger header during both water and air discharge through the sparger is studied. In addition, pressure oscillation occurring in the IRWST during air discharge through the sparger is studied for a better understanding of mechanisms of air discharge md an advanced evaluation technology of reactor safety. Understanding of flow behaviors occurring m the various types of pipes when POSRV is opened are also very important because those are very complex and may damage the structures of reactor coolant system. The principle of shock tube has been applied to analyze flow behaviors occurring in the piping system and several important phenomena which can be used for the evaluation of nuclear reactor safety has been obtained.

Identification of significant problems related to light water reactor piping systems

International Nuclear Information System (INIS)

1980-07-01

Work on the project was divided into three tasks. In Task 1, past surveys of LWR piping system problems and recent Licensee Event Report summaries are studied to identify the significant problems of LWR piping systems and the primary causes of these problems. Pipe cracking is identified as the most recurring problem and is mainly due to the vibration of pipes due to operating pump-pipe resonance, fluid-flow fluctuations, and vibration of pipe supports. Research relevant to the identified piping system problems is evaluated. Task 2 studies identify typical LWR piping systems and the current loads and load combinations used in the design of these systems. Definitions of loads are reviewed. In Task 3, a comparative study is carried out on the use of nonlinear analysis methods in the design of LWR piping systems. The study concludes that the current linear-elastic methods of analysis may not predict accurately the behavior of piping systems under seismic loads and may, under certain circumstances, result in nonconservative designs. Gaps at piping supports are found to have a significant effect on the response of the piping systems

Thermal fatigue evaluation of piping system Tee-connections

International Nuclear Information System (INIS)

Metzner, K.J.; Braillard, O.; Faidy, C.; Marcelles, I.; Solin, J.; Stumpfrock, L.

2004-01-01

Thermal fatigue is one significant long-term degradation mechanism nuclear power plants (NPP), in particular, as operating plants become older and life time extension activities have been initiated. In general, the common thermal fatigue issues are understood and controlled by plant instrumentation systems. However, incidents in some plants indicate that certain piping system Tees are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentation. The THERFAT project has been initiated to advance the accuracy and reliability of thermal fatigue load determination in engineering tools and research oriented approaches to outline a science based practical methodology for managing thermal fatigue risks in Tee-connections susceptible to high cyclic thermal fatigue. (orig.)

Application of tearing modulus stability concepts to nuclear piping. Final report

International Nuclear Information System (INIS)

Cotter, K.H.; Chang, H.Y.; Zahoor, A.

1982-02-01

The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK

The PEACE PIPE: Recycling nuclear weapons into a TRU storage/shipping container

International Nuclear Information System (INIS)

Floyd, D.; Edstrom, C.; Biddle, K.; Orlowski, R.; Geinitz, R.; Keenan, K.; Rivera, M.

1997-01-01

This paper describes results of a contract undertaken by the National Conversion Pilot Project (NCPP) at the Rocky Flats Environmental Technology Site (RFETS) to fabricate stainless steel ''pipe'' containers for use in certification testing at Sandia National Lab, Albuquerque to qualify the container for both storage of transuranic (TRU) waste at RFETS and other DOE sites and shipping of the waste to the Waste Isolation Pilot Project (WIPP). The paper includes a description of the nearly ten-fold increase in the amount of contained plutonium enabled by the product design, the preparation and use of former nuclear weapons facilities to fabricate the components, and the rigorous quality assurance and test procedures that were employed. It also describes how stainless steel nuclear weapons components can be converted into these pipe containers, a true ''swords into plowshare'' success story

A study on prediction of metal loss by flow-accelerated corrosion in the CANDU NPP secondary piping systems

International Nuclear Information System (INIS)

Shim, S. H.; Song, J. S.; Yoon, K. B.; Hwang, K. M.; Jin, T. E.; Lee, S. H.; Kim, W. S.

2001-01-01

Flow-Accelerated Corrosion(FAC) is a phenomenon that results in metal loss from piping, vessels, and equipment made of carbon steel. FAC occurs only under certain conditions of flow, chemistry, geometry, and material. Unfortunately, those conditions are in much of the high-energy piping in nuclear and fossil-fueled power plants. Also, for domestic NPP secondary pipings whose operating time become longer, more evidences of FAC have been reported. The authors are studying on FAC management using CHECWORKS, computer code developed by EPRI. This paper is on the prediction results of metal loss by FAC in the one of CANDU type NPP secondary piping systems

PE 100 pipe systems

CERN Document Server

Brömstrup, Heiner

2012-01-01

English translation of the 3rd edition ""Rohrsysteme aus PE 100"". Because of the considerably increased performance, pipe and pipe systems made from 100 enlarge the range of applications in the sectors of gas and water supply, sewage disposal, industrial pipeline construction and in the reconstruction and redevelopment of defective pipelines (relining). This book applies in particular to engineers, technicians and foremen working in the fields of supply, disposal and industry. Subject matters of the book are all practice-relevant questions regarding the construction, operation and maintenance

ELIMINATING CONSERVATISM IN THE PIPING SYSTEM ANALYSIS PROCESS THROUGH APPLICATION OF A SUITE OF LOCALLY APPROPRIATE SEISMIC INPUT MOTIONS

International Nuclear Information System (INIS)

Crawford, Anthony L.; Spears, Robert E.; Russell, Mark J.

2009-01-01

Seismic analysis is of great importance in the evaluation of nuclear systems due to the heavy influence such loading has on their designs. Current Department of Energy seismic analysis techniques for a nuclear safety-related piping system typically involve application of a single conservative seismic input applied to the entire system (1). A significant portion of this conservatism comes from the need to address the overlapping uncertainties in the seismic input and in the building response that transmits that input motion to the piping system. The approach presented in this paper addresses these two sources of uncertainty through the application of a suite of 32 input motions whose collective performance addresses the total uncertainty while each individual motion represents a single variation of it. It represents an extension of the soil-structure interaction analysis methodology of SEI/ASCE 43-05 (2) from the structure to individual piping components. Because this approach is computationally intensive, automation and other measures have been developed to make such an analysis efficient. These measures are detailed in this paper

Further considerations for damping in heavily insulated pipe systems

International Nuclear Information System (INIS)

Anderson, M.J.; Lindquist, M.R.; Severud, L.K.

1985-01-01

Over the past several years a body of test data has been accumulated which demonstrates that damping in small diameter heavily insulated pipe systems is much larger than presently recommended by Regulatory Guide 1.61. This data is generally based on pipe systems using a stand-off insulation design with a heater annulus. Additional tests have how been completed on similar pipe systems using a strap-on insulation design without the heater annulus. Results indicate some reduction in damping over the stand-off designs. Test data has also been obtained on a larger sixteen-inch diameter heavily insulated pipe system. Results of these two additional test series are presented. Revised damping values for seismic design of heavily insulated pipe systems are then recommended

Further considerations for damping in heavily insulated pipe systems

International Nuclear Information System (INIS)

Anderson, M.J.; Lindquist, M.R.; Severud, L.K.

1985-01-01

Over the past several years a body of test data has been accumulated which demonstrates that damping in small diameter heavily insulated pipe systems is much larger than presently recommended by Regulatory Code 1.61. This data is generally based on pipe systems using a stand-off insulation design with a heater annulus. Additional tests have now been completed on similar pipe systems using a strap-on insulation design without the heater annulus. Results indicate some reduction in damping over the stand-off designs. Test data has also been obtained on a larger sixteen-inch diameter heavily insulated pipe system. Results of these two additional test series are presented. Revised damping values for seismic design for heavily insulated pipe systems are then recommended

Acoustic leak detection in piping systems, 4

International Nuclear Information System (INIS)

Kitajima, Akira; Naohara, Nobuyuki; Aihara, Akihiko

1983-01-01

To monitor a high-pressure piping of nuclear power plants, a possibility of acoustic leak detection method has been experimentally studied in practical field tests and laboratory tests. Characteristics of background noise in field test and the results of experiment are summarized as follows: (1) The level of background noise in primary loop (PWR) was almost constant under actual plant operation. But it is possible that it rises at the condition of the pressure in primary loop. (2) Based on many experience of laboratory tests and practical field tests. The leak monitoring system for practical field was designed and developed. To improve the reliability, a judgment of leak on this system is used three factors of noise level, duration time of phenomena and frequency spectrum of noise signal emitted from the leak point. (author)

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

International Nuclear Information System (INIS)

Radu, V.

2009-01-01

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

Analysis of a piping system for requalification

International Nuclear Information System (INIS)

Hsieh, B.J.; Tang, Yu.

1992-01-01

This paper discusses the global stress analysis required for the seismic/structural requalification of a reactor secondary piping system in which minor defects (flaws) were discovered during a detailed inspection. The flaws in question consisted of weld imperfections. Specifically, it was necessary to establish that the stresses at the flawed sections did not exceed the allowables and that the fatigue life remained within acceptable limits. At the same time the piping system had to be qualified for higher earthquake loads than those used in the original design. To accomplish these objectives the nominal stress distributions in the piping system under the various loads (dead load, thermal load, wind load and seismic load) were determined. First a best estimate finite element model was developed and calculations were performed using the piping analysis modules of the ANSYS Computer Code. Parameter studies were then performed to assess the effect of physically reasonable variations in material, structural, and boundary condition characteristics. The nominal stresses and forces so determined, provided input for more detailed analyses of the flawed sections. Based on the reevaluation, the piping flaws were judged to be benign, i.e., the piping safety margins were acceptable inspite of the increased seismic demand. 13 refs

Seismic analysis of liquid metal reactor piping systems

International Nuclear Information System (INIS)

Wang, C.Y.

1987-01-01

To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

30 CFR 75.1905-1 - Diesel fuel piping systems.

Science.gov (United States)

2010-07-01

... facility. (g) Diesel fuel piping systems from the surface shall only be used to transport diesel fuel... storage facility. (h) The diesel fuel piping system must not be located in a borehole with electric power... entry as electric cables or power lines. Where it is necessary for piping systems to cross electric...

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

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

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)

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.)

Influence of plastic deformation on seismic response of piping

International Nuclear Information System (INIS)

Yao Yanping; Chen Yong; Lu Mingwan

2000-01-01

On the basis of a brief summary of linear elastic seismic analysis methods, the importance for consideration of plastic deformation during the dynamic response analysis of piping system is indicated. The present methods of considering plasticity and the disadvantages of these methods are discussed. And the authors point out that in order to reduce the conservatism of present codes and to put forward appropriate and realistic piping seismic design methods, the key is to understand the plastic dynamic failure mode for piping under seismic excitation and to calculate the inelastic energy dissipation. The analysis and evaluation are applicable to nuclear piping systems

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

Concepts for benchmark problem development for fracture mechanics application in safety evaluation of nuclear piping in subcreep service

International Nuclear Information System (INIS)

Reich, M.; Esztergar, E.P.; Erdogan, F.; Gray, T.G.F.; Spence, J.

1979-01-01

This report provides basic concepts and a review of the problem areas associated with the development of analytical and experimental programs for a systematic evaluation and comparison of the currently available fracture mechanics theories. The basis for such an evaluation is conceived as a series of benchmark problems which are accurately specified examples of geometry, loading, and environmental conditions, characteristic of large diameter thin wall piping systems in nuclear service. Starting from the simplest test coupons for cracked plate specimens, the program is to be designed in such a way that the range of validity and relative merit of the competing assessment methods can be evaluated and the results applied to increasingly more complex test configurations and ultimately to real piping systems. (Auth.)

Estimates of margins in ASME Code strength values for stainless steel nuclear piping

International Nuclear Information System (INIS)

Ware, A.G.

1995-01-01

The margins in the ASME Code stainless steel allowable stress values that can be attributed to the variations in material strength are evaluated for nuclear piping steels. Best-fit curves were calculated for the material test data that were used to determine allowable stress values for stainless steels in the ASME Code, supplemented by more recent data, to estimate the mean stresses. The mean yield stresses (on which the stainless steel S m values are based) from the test data are about 15 to 20% greater than the ASME Code yield stress values. The ASME Code yield stress values are estimated to approximately coincide with the 97% confidence limit from the test data. The mean and 97% confidence limit values can be used in the probabilistic risk assessments of nuclear piping

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

Comparison of elastic and inelastic seismic response of high temperature piping systems

International Nuclear Information System (INIS)

Thomas, F.M.; McCabe, S.L.; Liu, Y.

1994-01-01

A study of high temperature power piping systems is presented. The response of the piping systems is determined when subjected to seismic disturbances. Two piping systems are presented, a main steam line, and a cold reheat line. Each of the piping systems are modeled using the ANSYS computer program and two analyses are performed on each piping system. First, each piping system is subjected to a seismic disturbance and the pipe material is assumed to remain linear and elastic. Next the analysis is repeated for each piping system when the pipe material is modeled as having elastic-plastic behavior. The results of the linear elastic analysis and elastic-plastic analysis are compared for each of the two pipe models. The pipe stresses, strains, and displacements, are compared. These comparisons are made so that the effect of the material yielding can be determined and to access what error is made when a linear analysis is performed on a system that yields

Dynamic response of piping system subject to flow acoustic excitation

International Nuclear Information System (INIS)

Wang, T.; Sun, Y.S.

1988-01-01

Through the use of a theoretically derived and test data-calibrated forcing function, the dynamic response of a piping system subject to flow-acoustic induced vibration is analyzed. It is shown that the piping behavior can be predicted when consideration is given to both the wall flexural vibration and the piping system vibration. Piping responded as a system to the transversal excitation due to the swirling motion of the fluid flow, as well as flexurally to the high-frequency acoustic excitations. The transverse piping system response was calculated using a lumped mass piping model. The piping model has more stringent requirements than its counterpart for waterhammer and seismic modeling due to the shorter spiral wavelength and higher frequency of the forcing function. Proper modeling ensured that both the moment stress caused by system excitation and the local stress induced by the support reaction load were properly accounted for. Flexural vibration not only poses a threat to nipples and branch connections, but also contributes substantially to the resultant total stress experienced by the pipe. The forcing function approach has the advantage that the critical locations on the piping system can be identified by means of analysis, facilitating surveillance and inspection, as well as fatigue evaluation

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

International Nuclear Information System (INIS)

Oliveira Segabinaze, R. de

1994-01-01

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

FRAMEWORK FOR STRUCTURAL ONLINE HEALTH MONITORING OF AGING AND DEGRADATION OF SECONDARY PIPING SYSTEMS DUE TO SOME ASPECTS OF EROSION

Energy Technology Data Exchange (ETDEWEB)

Gribok, Andrei V.; Agarwal, Vivek

2017-06-01

This paper describes the current state of research related to critical aspects of erosion and selected aspects of degradation of secondary components in nuclear power plants (NPPs). The paper also proposes a framework for online health monitoring of aging and degradation of secondary components. The framework consists of an integrated multi-sensor modality system, which can be used to monitor different piping configurations under different degradation conditions. The report analyses the currently known degradation mechanisms and available predictive models. Based on this analysis, the structural health monitoring framework is proposed. The Light Water Reactor Sustainability Program began to evaluate technologies that could be used to perform online monitoring of piping and other secondary system structural components in commercial NPPs. These online monitoring systems have the potential to identify when a more detailed inspection is needed using real time measurements, rather than at a pre-determined inspection interval. This transition to condition-based, risk-informed automated maintenance will contribute to a significant reduction of operations and maintenance costs that account for the majority of nuclear power generation costs. Furthermore, of the operations and maintenance costs in U.S. plants, approximately 80% are labor costs. To address the issue of rising operating costs and economic viability, in 2017, companies that operate the national nuclear energy fleet started the Delivering the Nuclear Promise Initiative, which is a 3 year program aimed at maintaining operational focus, increasing value, and improving efficiency. There is unanimous agreement between industry experts and academic researchers that identifying and prioritizing inspection locations in secondary piping systems (for example, in raw water piping or diesel piping) would eliminate many excessive in-service inspections. The proposed structural health monitoring framework takes aim at

Failure rates in piping manufactured to different standards

International Nuclear Information System (INIS)

Barnes, R.W.; Cooper, G.D.

1995-11-01

Most non-nuclear process piping systems in Canada and the United States are constructed to the requirements of the piping codes of the American Society of Mechanical Engineers (ASME B31.1 and B31.3). Section III of the ASME Boiler and Pressure Vessel Code, has additional requirements for piping that are expected to provide further assurance of pressure boundary integrity. This project attempted to determine if the additional requirements of Section III were beneficial in preventing failure of the pressure boundary. The approach taken in the study was to determine the causes of failure of non-nuclear piping subjected to service similar to that experienced by piping in CANDU nuclear power plants. The study examined information on carbon steel piping systems filled with water/steam which operate up to a maximum temperature of 600 F and a maximum pressure of 1600 psi. The failure mechanisms were identified and analysed to determine whether application of the requirements of Section III would have prevented the failure. Through a process of interviews and literature search, 186 failures were identified and assembled into a reference database. Many of the records were incomplete; therefore, the reference database was trimmed to include a subset of 65 failure points supported by complete data. This subset formed the basis for this study. The results from the study of other databases assembled for similar purposes were reviewed and compared to the conclusions reached in this study. These reviews confirmed the conclusions reached in this study. (author). 48 refs., 20 tabs

Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

International Nuclear Information System (INIS)

Rahardjo, H.P.

2011-01-01

Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipment. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogeneously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occurred in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code) was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs) of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA) in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, expansion load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (re-routing) is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node 30, also a

Pipe/duct system design for tornado missile impact loads

Energy Technology Data Exchange (ETDEWEB)

Li, J.; Wang, S.; Johnson, W., E-mail: whjohnso@bechtel.com

2014-04-01

For nuclear power plant life extension projects, it may be convenient and in some instances necessary to locate safety-related steel ducts and pipes outside of the main structures, exposing them to extreme environmental loads such as tornado missile impact. Examples of this application include emergency firewater lines and Control Room vent ducts. A typical exposed commodity run could be comprised of a rectangular or circular cross-section with horizontal and vertical segments supported at variable spans off of roof and wall panels, respectively. Efficient and economical design of such a tornado-impacted duct or pipe system, consisting of the commodity and its supports, must exploit all of the system's capability to absorb the impact energy by deforming plastically to the fullest extent allowable. Energy can be absorbed locally in the vicinity of impact on the commodity, globally through rotation at flexural plastic hinges, and through yielding of the supports. In this paper a simplified NDOF lumped parameter nonlinear analysis methodology is presented and applied to the coupled commodity/support system subjected to tornado impulse loading. The analysis methodology is confirmed using a detailed ANSYS nonlinear finite element model. Optimization of the initial trial design is achieved by progressively decreasing the support resistances, while monitoring the response ductilities throughout the system. Evaluation methodologies are provided for the four types of plastic deformation responses which occur in the system: local response in the immediate vicinity of impact, flexural and membrane response of the sidewall out to one or two times the commodity depth beyond the point of impact, global response of the commodity as a beam spanning between supports, and the shear and flexural response of support. The inelastic responses are evaluated against AISC N690 acceptance criteria (ANSI, 2006), supplemented as appropriate by triaxiality considerations for inelastic

Seismic testing and analysis of a prototypic nonlinear piping system

International Nuclear Information System (INIS)

Barta, D.A.; Anderson, M.J.; Severud, L.K.

1982-11-01

A series of seismic tests and analyses of a nonlinear Fast Flux Test Facility (FFTF) prototypic piping system are described, and measured responses are compared with analytical predictions. The test loop was representative of a typical LMFBR insulated small bore piping system and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps. Various piping support configurations were tested and analyzed to evaluate the effects of free play and other nonlinear stiffness characteristics on the piping system response

Response of HDR-VKL piping system to seismic test excitations: Comparison of analytical predictions and test measurements

International Nuclear Information System (INIS)

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

1989-01-01

As part of the earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, simulated seismic tests (SHAM) were performed during April--May 1988 on the VKL (Versuchskreislauf) piping system. The purpose of these experiments was to study the behavior of piping subjected to a range of seismic excitation levels including those that exceed design levels manifold and that might induce failure of pipe supports or plasticity in the pipe runs, and to establish seismic margins for piping and pipe supports. Data obtained in the tests are also used to validate analysis methods. Detailed reports on the SHAM experiments are given elsewhere. The objective of this document is to evaluate a subsystem analysis module of the SMACS code. This module is a linear finite-element based program capable of calculating the response of nuclear power plant subsystems subjected to independent multiple-acceleration input excitation. The evaluation is based on a comparison of computational results of simulation of SHAM tests with corresponding test measurements

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

Seismic proving test of ultimate piping strength (current status of preliminary tests)

International Nuclear Information System (INIS)

Suzuki, K.; Namita, Y.; Abe, H.; Ichihashi, I.; Suzuki, K.; Ishiwata, M.; Fujiwaka, T.; Yokota, H.

2001-01-01

In 1998 Fiscal Year, the 6 year program of piping tests was initiated with the following objectives: i) to clarify the elasto-plastic response and ultimate strength of nuclear piping, ii) to ascertain the seismic safety margin of the current seismic design code for piping, and iii) to assess new allowable stress rules. In order to resolve extensive technical issues before proceeding on to the seismic proving test of a large-scale piping system, a series of preliminary tests of materials, piping components and simplified piping systems is intended. In this paper, the current status of the material tests and the piping component tests is reported. (author)

Calculation of the pipes failure probability of the Rcic system of a nuclear power station by means of software WinPRAISE 07; Calculo de la probabilidad de falla de tuberias del sistema RCIC de una central nuclear mediante el software WinPRAISE 07

Energy Technology Data Exchange (ETDEWEB)

Jasso G, J.; Diaz S, A.; Mendoza G, G.; Sainz M, E. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Garcia de la C, F. M., E-mail: angeles.diaz@inin.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

2014-10-15

The growth and the cracks propagation by fatigue are a typical degradation mechanism that is presented in the nuclear industry as in the conventional industry; the unstable propagation of a crack can cause the catastrophic failure of a metallic component even with high ductility; for this reason, activities of programmed maintenance have been established in the industry using inspection and visual techniques and/or ultrasound with an established periodicity allowing to follow up to these growths, controlling the undesirable effects; however, these activities increase the operation costs; and in the peculiar case of the nuclear industry, they increase the radiation exposure to the participant personnel. The use of mathematical processes that integrate concepts of uncertainty, material properties and the probability associated to the inspection results, has been constituted as a powerful tool of evaluation of the component reliability, reducing costs and exposure levels. In this work the evaluation of the failure probability by cracks growth preexisting by fatigue is presented, in pipes of a Reactor Core Isolation Cooling system (Rcic) in a nuclear power station. The software WinPRAISE 07 (Piping Reliability Analysis Including Seismic Events) was used supported in the probabilistic fracture mechanics principles. The obtained values of failure probability evidenced a good behavior of the analyzed pipes with a maximum order of 1.0 E-6, therefore is concluded that the performance of the lines of these pipes is reliable even extrapolating the calculations at 10, 20, 30 and 40 years of service. (Author)

Using data visualization tools to support degradation assessment in nuclear piping

International Nuclear Information System (INIS)

Jyrkama, M.I.; Pandey, M.D.

2012-01-01

Nuclear utilities collect a vast amount of in-service inspection data as part of periodic inspection plans and the detailed assessment and monitoring of various degradation mechanisms, such as fretting, corrosion, and creep. In many cases, the focus is primarily on ensuring that the observed minimum or maximum values are within the acceptable regulatory limits, while the rest of the (often costly) surveillance data remains unused and unanalyzed. The objective of this study is to illustrate how data visualization tools can be used effectively to analyze and consider all of the in-service inspection data, and hence provide valuable support for the degradation assessment in nuclear piping. The 2D and 3D visualization tools discussed in this paper were developed mainly in the context of flow accelerated corrosion (FAC) assessment in feeder piping, where the complex pipe geometries and flow conditions have a significant impact on the ultrasonic (UT) wall thickness measurements. The visualization of eddy current inspection results from the assessment of pitting corrosion of steam generator tubing will also be discussed briefly. The visualization tools provide a more comprehensive view of the degree and extent of degradation, and hence directly support the planning of future inspection of critical components by identifying key locations and areas for detailed monitoring. The results furthermore increase the confidence and reliability of fitness-for-service (FFS) assessments and life cycle management (LCM) planning decisions with respect to component repair or replacement. (author)

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.

Piping data retrieval system (PDRS): An integrated package to aid piping layout

International Nuclear Information System (INIS)

Vyas, K.N.; Sharma, A.; Susandhi, R.; Basu, S.

1986-01-01

An integrated package to aid piping layout has been developed and implemented on PDP-11/34 system at Hall 7. The package allows various equipments to be modelled, consisting of primitive equipment components. The equipment layout for the plant can then be reproduced in the form of drawings such as plan, elevation, isometric or perspective. The package has the built in function to perform hidden line removal among equipments. Once the equipment layout is finalised, the package aids in superimposing the piping as per the specified pipe routine. The report discusses the general capabilities and the major input requirements for the package. (author)

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

Pressure piping systems examination. 2. ed

Energy Technology Data Exchange (ETDEWEB)

1993-03-01

This Code is Part 13 of the IP Model Code of Safe Practice in the Petroleum Industry. Its purpose is to provide a guide to safe practices in the in-service examination and test of piping systems used in the petroleum and chemical industries. The Code gives general requirements regarding the provision and maintenance of adequate documentation, in-service examination, the control of modifications and repairs, examination frequency, protective devices and testing of piping systems. (author)

Status report on nuclear electric propulsion systems

Science.gov (United States)

Stearns, J. W.

1975-01-01

Progress in nuclear electric propulsion (NEP) systems for a multipayload multimission vehicle needed in both deep-space missions and a variety of geocentric missions is reviewed. The space system power level is a function of the initial launch vehicle mass, but developments in out-of-core nuclear thermionic direct conversion have broadened design options. Cost, design, and performance parameters are compared for reusable chemical space tugs and NEP reusable space tugs. Improvements in heat pipes, ion engines, and magnetoplasmadynamic arc jet thrust subsystems are discussed.

View of industry on the impact of pipe break criteria

International Nuclear Information System (INIS)

Bernsen, S.A.

1983-01-01

Historically, large pipe breaks in the types of materials used and under operating conditions similar to those in light water reactor service have not occurred. Nevertheless, the non-mechanistic assumption of a double ended pipe break of the early sixties, selected for loss of coolant accident analysis purposes, has become a mechanistic criterion for the design and arrangement of high pressure piping systems and their associated supports and enclosures in today's nuclear plants. While it seems reasonable and appropriate to continue to design the Emergency Core Cooling Systems for a range of loss of coolant accidents up to and including those that approximate the area of the largest pipe connected to the reactor vessel and to use this break in determining the loading and temperature rise rate for containment structures and equipment qualification, it no longer seems reasonable to provide precisely engineered break protection for a limited number of potential pipe break locations. This observation is gaining increasing support, particularly as engineering judgment and historical perspectives are being supplemented by both deterministic and probabilistic studies that indicate the potential for large instantaneous breaks in nuclear grade piping systems is virtually incredible. Fracture mechanics analyses support leak-before-break assumptions with wide margins and probabilistic studies indicate potentials for double-ended pipe breaks in the range of less than one in a billion years

BOA: Pipe-asbestos insulation removal robot system

International Nuclear Information System (INIS)

Schempf, H.; Bares, J.; Mutschler, E.

1995-01-01

This paper describes the BOA system, a mobile pipe-external crawler used to remotely strip and bag (possibly contaminated) asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations across the DOE weapons complex. The mechanical removal of ACLIM is very cost-effective due to the relatively low productivity and high cost involved in human removal scenarios. BOA, a mechanical system capable of removing most forms of lagging (paper, plaster, aluminum sheet, clamps, screws and chicken-wire), and insulation (paper, tar, asbestos fiber, mag-block) uses a circular cutter and compression paddles to cut and strip the insulation off the pipe through compression, while a HEPA-filter and encapsulant system maintain a certifiable vacuum and moisture content inside the system and on the pipe, respectively. The crawler system has been built and is currently undergoing testing. Key design parameters and performance parameters are developed and used in performance testing. Since the current system is a testbed, we also discuss future enhancements and outline two deployment scenarios (robotic and manual) for the final system to be designed and completed by the end of FY '95. An on-site demonstration is currently planned for Fernald in Ohio and Oak Ridge in Tennessee

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)

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

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)

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)

Pipe Explorer{trademark} surveying system. Innovative technology summary report

Energy Technology Data Exchange (ETDEWEB)

1999-06-01

The US Department of Energy`s (DOE) Chicago Operations Office and the DOE`s Federal Energy Technology Center (FETC) developed a Large Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial decontamination and decommissioning (D and D) technologies in comparison with current baseline technologies. The Pipe Explorer{trademark} system was developed by Science and Engineering Associates, Inc. (SEA), Albuquerque, NM as a deployment method for transporting a variety of survey tools into pipes and ducts. Tools available for use with the system include alpha, beta and gamma radiation detectors; video cameras; and pipe locator beacons. Different versions of this technology have been demonstrated at three other sites; results of these demonstrations are provided in an earlier Innovative Technology Summary Report. As part of a D and D project, characterization radiological contamination inside piping systems is necessary before pipes can be recycled, remediated or disposed. This is usually done manually by surveying over the outside of the piping only, with limited effectiveness and risk of worker exposure. The pipe must be accessible to workers, and embedded pipes in concrete or in the ground would have to be excavated at high cost and risk of exposure to workers. The advantage of the Pipe Explorer is its ability to perform in-situ characterization of pipe internals.

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

Effect of piping systems on surge in centrifugal compressors

International Nuclear Information System (INIS)

Tamaki, Hideaki

2008-01-01

There is a possibility that the exchange of the piping system may change the surge characteristic of a compressor. The piping system of a plant is not always the same as that of a test site. Then it is important to evaluate the effect of piping systems on surge characteristics in centrifugal compressors. Several turbochargers combined with different piping systems were tested. The lumped parameter model which was simplified to be solved easily was applied for the prediction of surge point. Surge lines were calculated with the linearlized lumped parameter model. The difference between the test and calculated results was within 10 %. Trajectory of surge cycle was also examined by solving the lumped parameter model. Mild surge and deep surge were successfully predicted. This study confirmed that the lumped parameter model was a very useful tool to predict the effect of piping systems on surge characteristics in centrifugal compressors, even though that was a simple model

Analysis and computer simulation for transient flow in complex system of liquid piping

International Nuclear Information System (INIS)

Mitry, A.M.

1985-01-01

This paper is concerned with unsteady state analysis and development of a digital computer program, FLUTRAN, that performs a simulation of transient flow behavior in a complex system of liquid piping. The program calculates pressure and flow transients in the liquid filled piping system. The analytical model is based on the method of characteristics solution to the fluid hammer continuity and momentum equations. The equations are subject to wide variety of boundary conditions to take into account the effect of hydraulic devices. Water column separation is treated as a boundary condition with known head. Experimental tests are presented that exhibit transients induced by pump failure and valve closure in the McGuire Nuclear Station Low Level Intake Cooling Water System. Numerical simulation is conducted to compare theory with test data. Analytical and test data are shown to be in good agreement and provide validation of the model

Optimal support arrangement of piping systems using genetic algorithm

International Nuclear Information System (INIS)

Chiba, T.; Okado, S.; Fujii, I.; Itami, K.

1996-01-01

The support arrangement is one of the important factors in the design of piping systems. Much time is required to decide the arrangement of the supports. The authors applied a genetic algorithm to find the optimum support arrangement for piping systems. Examples are provided to illustrate the effectiveness of the genetic algorithm. Good results are obtained when applying the genetic algorithm to the actual designing of the piping system

CAPD Software Development for Automatic Piping System Design: Checking Piping Pocket, Checking Valve Level and Flexibility

International Nuclear Information System (INIS)

Ari Satmoko; Edi Karyanta; Dedy Haryanto; Abdul Hafid; Sudarno; Kussigit Santosa; Pinitoyo, A.; Demon Handoyo

2003-01-01

One of several steps in industrial plant construction is preparing piping layout drawing. In this drawing, pipe and all other pieces such as instrumentation, equipment, structure should be modeled A software called CAPD was developed to replace and to behave as piping drafter or designer. CAPD was successfully developed by adding both subprogram CHKUPIPE and CHKMANV. The first subprogram can check and gives warning if there is piping pocket in the piping system. The second can identify valve position and then check whether valve can be handled by operator hand The main program CAPD was also successfully modified in order to be capable in limiting the maximum length of straight pipe. By limiting the length, piping flexibility can be increased. (author)

Comparison of ICEPEL predictions with single elbow flexible piping system experiment

International Nuclear Information System (INIS)

A-Moneim, M.T.; Chang, Y.W.

1978-01-01

The ICEPEL Code for coupled hydrodynamic-structural response analysis of piping systems is used to analyze an experiment on the response of flexible piping systems to internal pressure pulses. The piping system consisted of two flexible Nickel-200 pipes connected in series through a 90 0 thick-walled stainless steel elbow. A tailored pressure pulse generated by a calibrated pulse gun is stabilized in a long thick-walled stainless steel pipe leading to the flexible piping system which ended with a heavy blind flange. The analytical results of pressure and circumferential strain histories are discussed and compared against the experimental data obtained by Stanford Research Institute

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)

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

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

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)

Pipe connector

International Nuclear Information System (INIS)

Sullivan, T.E.; Pardini, J.A.

1978-01-01

A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated

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

International Nuclear Information System (INIS)

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

1982-08-01

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

Study of a risk-based piping inspection guideline system.

Science.gov (United States)

Tien, Shiaw-Wen; Hwang, Wen-Tsung; Tsai, Chih-Hung

2007-02-01

A risk-based inspection system and a piping inspection guideline model were developed in this study. The research procedure consists of two parts--the building of a risk-based inspection model for piping and the construction of a risk-based piping inspection guideline model. Field visits at the plant were conducted to develop the risk-based inspection and strategic analysis system. A knowledge-based model had been built in accordance with international standards and local government regulations, and the rational unified process was applied for reducing the discrepancy in the development of the models. The models had been designed to analyze damage factors, damage models, and potential damage positions of piping in the petrochemical plants. The purpose of this study was to provide inspection-related personnel with the optimal planning tools for piping inspections, hence, to enable effective predictions of potential piping risks and to enhance the better degree of safety in plant operations that the petrochemical industries can be expected to achieve. A risk analysis was conducted on the piping system of a petrochemical plant. The outcome indicated that most of the risks resulted from a small number of pipelines.

Crack initiation through vibration fatigue of small-diameter pipes

International Nuclear Information System (INIS)

Comby, R.; Thebault, Y.; Papaconstantinou, T.

2002-01-01

Socket welds are used extensively for small bore piping connections in nuclear power plant systems. Numerous fatigue-related failures occurred in the past ten years mainly on safeguard systems and continue to occur frequently, showing that corrective actions did not take into account all aspects of the problem. Destructive examination of cracked small bore piping connections allowed a better understanding of failure mechanisms and a prediction of crack initiation site depending on nozzle fittings such as run pipe and small bore pipe thickness. A three-dimensional finite element model confirmed the conclusions of the lab examinations. For thick run pipes, it was shown that the failure tend to initiate predominantly at the socket weld toe or at the root, depending on the respective thickness of coupling and small bore pipe. Some additional studies, based on RSE-M code, are in progress in order to determine the maximum stresses location. Lessons learned through these investigations led to optimise the in-service inspection scope and to define solutions to be carried out to prevent failure of ''susceptible'' small bore pipe connections. Since July 2000, a large program is in progress to select all ''susceptible'' small bore pipes in safety-related systems and to apply corrective measures such as piping modifications or system operational modifications. (authors)

Surveys of embedded piping for Shoreham license termination

International Nuclear Information System (INIS)

Williams, D.E. Jr.

2004-01-01

In planning the decommissioning of the Shoreham Nuclear Power Station (SNPS) in Wading River, N.Y., it was determined that the cost of removing contaminated floor drain piping was prohibitive. The piping is typically embedded approximately four feet deep in reinforced concrete, often below structural I-beams. A decision was made to develop remote survey devices ('pipe crawlers') that would allow SNPS to decontaminate and survey embedded piping within NRC free release limits. Pipe crawlers currently in use at SNPS are able to traverse multiple 45 and 90 degree bends while maintaining all detectors in the required geometry (less than 1 cm detector to surface distance). The following aspects of this project will be presented: 1) System classification and cost-benefit analysis 2) Overview of system decontamination 3) Pipe crawler mechanical and electrical development 4) Detector backgrounds and MDA's 5) Additional devices and techniques 6) NRC position on crawler use. 7) SNPS results to date. (author)

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

International Nuclear Information System (INIS)

Mohan, R.; Marschall, C.; Krishnaswamy, P.; Brust, F.; Ghadiali, N.; Wilkowski, G.

1995-04-01

This topical report summarizes the work on angled crack growth and combined loading effects performed within the Nuclear Regulatory Commission's research program entitled open-quotes Short Cracks in Piping and Piping Weldsclose quotes. The major impetus for this work stemmed from the observation that initial circumferential cracks in carbon steel pipes exhibited angular crack growth. This failure mode was little understood, and the effect of angled crack growth from an initially circumferential crack raised questions of how pipes under combined loading with torsional stresses would behave. There were three major conclusions from this work. The first was that virtually all ferritic nuclear pipes will have toughness anisotropy. The second was that the ratio of the normalized crack driving force (as a function of angle) to the normalized toughness (also as a function of the angle of crack growth) showed that there was an equal likelihood of cracks growing at any angle between 25 and 65 degrees. This agreed with the scatter of crack growth angles observed in pipe tests. Third, for combined loads with torsional stresses, an effective moment allows pure bending analyses to be used up to crack initiation. Crack opening area under combined loads could also be determined in this mariner

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

Energy Technology Data Exchange (ETDEWEB)

Mohan, R.; Marschall, C.; Krishnaswamy, P.; Brust, F.; Ghadiali, N.; Wilkowski, G. [Battelle, Columbus, OH (United States)

1995-04-01

This topical report summarizes the work on angled crack growth and combined loading effects performed within the Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks in Piping and Piping Welds{close_quotes}. The major impetus for this work stemmed from the observation that initial circumferential cracks in carbon steel pipes exhibited angular crack growth. This failure mode was little understood, and the effect of angled crack growth from an initially circumferential crack raised questions of how pipes under combined loading with torsional stresses would behave. There were three major conclusions from this work. The first was that virtually all ferritic nuclear pipes will have toughness anisotropy. The second was that the ratio of the normalized crack driving force (as a function of angle) to the normalized toughness (also as a function of the angle of crack growth) showed that there was an equal likelihood of cracks growing at any angle between 25 and 65 degrees. This agreed with the scatter of crack growth angles observed in pipe tests. Third, for combined loads with torsional stresses, an effective moment allows pure bending analyses to be used up to crack initiation. Crack opening area under combined loads could also be determined in this mariner.

Seismic analysis response factors and design margins of piping systems

International Nuclear Information System (INIS)

Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

1985-01-01

The objective of the simplified methods project of the Seismic Safety Margins Research Program is to develop a simplified seismic risk methodology for general use. The goal is to reduce seismic PRA costs to roughly 60 man-months over a 6 to 8 month period, without compromising the quality of the product. To achieve the goal, it is necessary to simplify the calculational procedure of the seismic response. The response factor approach serves this purpose. The response factor relates the median level response to the design data. Through a literature survey, we identified the various seismic analysis methods adopted in the U.S. nuclear industry for the piping system. A series of seismic response calculations was performed. The response factors and their variabilities for each method of analysis were computed. A sensitivity study of the effect of piping damping, in-structure response spectra envelop method, and analysis method was conducted. In addition, design margins, which relate the best-estimate response to the design data, are also presented

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

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

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)

Incremental-hinge piping analysis methods for inelastic seismic response prediction

International Nuclear Information System (INIS)

Jaquay, K.R.; Castle, W.R.; Larson, J.E.

1989-01-01

This paper proposes nonlinear seismic response prediction methods for nuclear piping systems based on simplified plastic hinge analyses. The simplified plastic hinge analyses utilize an incremental series of flat response spectrum loadings and replace yielded components with hinge elements when a predefined hinge moment is reached. These hinge moment values, developed by Rodabaugh, result in inelastic energy dissipation of the same magnitude as observed in seismic tests of piping components. Two definitions of design level equivalent loads are employed: one conservatively based on the peaks of the design acceleration response spectra, the other based on inelastic frequencies determined by the method of Krylov and Bogolyuboff recently extended by Lazzeri to piping. Both definitions account for piping system inelastic energy dissipation using Newmark-Hall inelastic response spectrum reduction factors and the displacement ductility results of the incremental-hinge analysis. Two ratchet-fatigue damage models are used: one developed by Rodabaugh that conservatively correlates Markl static fatigue expressions to seismic tests to failure of piping components; the other developed by Severud that uses the ratchet expression of Bree for elbows and Edmunds and Beer for straights, and defines ratchet-fatigue interaction using Coffin's ductility based fatigue equation. Comparisons of predicted behavior versus experimental results are provided for a high-level seismic test of a segment of a representative nuclear plant piping system. (orig.)

A contribution for stress analysis in bend acessories of piping systems

International Nuclear Information System (INIS)

Melo, F.J.M.Q. de; Castro, P.M.S.T. de

1986-01-01

Analytical and numerical studies of the linear elastic behavior of bend pipes, with tangent pipes or flanged ends, such as used in nuclear power plants are presented. Two analytical techniques were developed; one is based on the integration of Euler equation and the other one is based on a Fourier analysis. The results obtained using these approaches are compared with results obtained by a finite element code for 'semiloof shells. (Author) [pt

Design of megawatt power level heat pipe reactors

Energy Technology Data Exchange (ETDEWEB)

Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-12

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

The development of a diagnostic and monitoring system for vibrating pipe branches

International Nuclear Information System (INIS)

Tanaka, Toshio; Suzuki, Shigeki; Nekomoto, Yoshitsugu; Tanaka, Mamoru

1994-01-01

This paper deals with a diagnostic and monitoring system for assessing the integrity of pipe branches, during the operation of the nuclear power plant. This system have been developed under the concept of ''easy to use without any sophisticated analysis'' and ''portable''. The accuracy of the diagnosis is based on the model optimization subsystem, which automatically modifies the numerical vibration model so as to fit its natural frequency to the actual natural frequency. The information obtained by this system may be reflected to a maintenance program of the plant to assure more reliable operation of the plant. (orig.)

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

Investigation on field method using strain measurement on pipe surface to measure pressure pulsation in piping systems