Leak-before-break behaviour of nuclear piping systems

International Nuclear Information System (INIS)

Bartholome, G.; Wellein, R.

1992-01-01

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

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)

Effect of pipe rupture loads inside containment in the break exclusionary piping outside containment

International Nuclear Information System (INIS)

Weiss, G.

1987-01-01

The plant design for protection against piping failures outside containment should make sure that fluid system piping in containment penetration areas are designed to meet the break exclusionary provisions contained in the BTP MEB 3-1. According to these provisions, following a piping failure (main steam line) inside containment, the part of the flued head connected to the piping outside containment, should not exceed the ASME Code stress limits for the appropriate load combinations. A finite element analysis has been performed to evaluate the stress level in this area. (orig./HP)

Probabilistic analyses of failure in reactor coolant piping

International Nuclear Information System (INIS)

Holman, G.S.

1984-01-01

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

Study on criterion for leak before break assessment of pressure pipes

International Nuclear Information System (INIS)

Yang Linjuan

2009-01-01

Based on the elastoplastic fracture mechanics, this paper established the expression formulas of limit buckling pressure P u on the ligament of axial and circumferential surface cracks and the initial pressure for the through cracks P c . A new Leak Before Break (LBB) assessment criterion was put forward to predict the failure mode of pressure pipes, i.e., when P u is less than P c , the pipe will leak; when P u is equal to or larger than P c , the pipe will break, which is verified by the test data reported in literatures. (authors)

Probabilistic fracture mechanics analysis for leak-before-break evaluation of light water reactor's piping

International Nuclear Information System (INIS)

Yoshimura, Shinobu; Yagawa, Genki; Akiba, Hiroshi; Fujioka, Terutaka.

1997-01-01

This paper describes Probabilistic Fracture Mechanics (PFM) analyses for quantitative evaluation of the likelihood of Leak-Before-Break (LBB) of Light Water Reactor's (LWR's) piping. The PFM analyses in general assume probabilistic distributions of initial crack size, applied stress cycles, crack growth laws, fracture criteria, leakage detection capability, defect inspection capability and so on. Referring to the deterministic procedure for LBB evaluation, most appropriate PFM models and data for LBB evaluation are discussed. Here the LBB index is newly proposed in order to quantitatively evaluate the likelihood of LBB. Through intensive sensitivity analyses, it is clarified that the LBB is more likely to occur for larger diameter pipe; the performance of leakage detection significantly affects the LBB likelihood; the LBB likelihood increases with plant's aging even conservatively assuming leak detection capability; the R6 method (Category 1, Option 1) for fracture criterion gives very conservative results; and In-Service Inspection (ISI) reduces the increase rate of failure probability than the failure probability itself. (author)

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

Statistical Dependence of Pipe Breaks on Explanatory Variables

Directory of Open Access Journals (Sweden)

Patricia Gómez-Martínez

2017-02-01

Full Text Available Aging infrastructure is the main challenge currently faced by water suppliers. Estimation of assets lifetime requires reliable criteria to plan assets repair and renewal strategies. To do so, pipe break prediction is one of the most important inputs. This paper analyzes the statistical dependence of pipe breaks on explanatory variables, determining their optimal combination and quantifying their influence on failure prediction accuracy. A large set of registered data from Madrid water supply network, managed by Canal de Isabel II, has been filtered, classified and studied. Several statistical Bayesian models have been built and validated from the available information with a technique that combines reference periods of time as well as geographical location. Statistical models of increasing complexity are built from zero up to five explanatory variables following two approaches: a set of independent variables or a combination of two joint variables plus an additional number of independent variables. With the aim of finding the variable combination that provides the most accurate prediction, models are compared following an objective validation procedure based on the model skill to predict the number of pipe breaks in a large set of geographical locations. As expected, model performance improves as the number of explanatory variables increases. However, the rate of improvement is not constant. Performance metrics improve significantly up to three variables, but the tendency is softened for higher order models, especially in trunk mains where performance is reduced. Slight differences are found between trunk mains and distribution lines when selecting the most influent variables and models.

Uncertainty analysis for probabilistic pipe fracture evaluations in LBB applications

International Nuclear Information System (INIS)

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

1997-01-01

During the NRC's Short Cracks in Piping and Piping Welds Program at Battelle, a probabilistic methodology was developed to conduct fracture evaluations of circumferentially cracked pipes for application to leak-rate detection. Later, in the IPIRG-2 program, several parameters that may affect leak-before-break and other pipe flaw evaluations were identified. This paper presents new results from several uncertainty analyses to evaluate the effects of normal operating stresses, normal plus safe-shutdown earthquake stresses, off-centered cracks, restraint of pressure-induced bending, and dynamic and cyclic loading rates on the conditional failure probability of pipes. systems in BWR and PWR. For each parameter, the sensitivity to conditional probability of failure and hence, its importance on probabilistic leak-before-break evaluations were determined

Uncertainty analysis for probabilistic pipe fracture evaluations in LBB applications

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

During the NRC`s Short Cracks in Piping and Piping Welds Program at Battelle, a probabilistic methodology was developed to conduct fracture evaluations of circumferentially cracked pipes for application to leak-rate detection. Later, in the IPIRG-2 program, several parameters that may affect leak-before-break and other pipe flaw evaluations were identified. This paper presents new results from several uncertainty analyses to evaluate the effects of normal operating stresses, normal plus safe-shutdown earthquake stresses, off-centered cracks, restraint of pressure-induced bending, and dynamic and cyclic loading rates on the conditional failure probability of pipes. systems in BWR and PWR. For each parameter, the sensitivity to conditional probability of failure and hence, its importance on probabilistic leak-before-break evaluations were determined.

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)

Effects of blast wave to main steam piping under high energy line break condition by TNT model

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Hyun; Lee, Eung Seok; Chang, Yoon Suk [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The aim of this study is to examine effect of the blast wave according to pipe break position through FE (Finite Element) analyses. If HELB (High Energy Line Break) accident occurs in nuclear power plants, not only environmental effect such as release of radioactive material but also secondary structural defects should be considered. Sudden pipe rupture causes ejection of high temperature and pressure fluid, which acts as a blast wave around the break location. The blast wave caused by the HELB has a possibility to induce structural defects around the components such as safe-related injection pipes and other structures.

Modeling of pipe break accident in a district heating system using RELAP5 computer code

International Nuclear Information System (INIS)

Kaliatka, A.; Valinčius, M.

2012-01-01

Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

Structural integrity of whipping pipes following a postulated circumferential break - a contribution to determining strain levels acceptable under faulted conditions

International Nuclear Information System (INIS)

Charalambus, B.; Labes, M.

1993-01-01

It is postulated that a break of a thin-walled pipe does not cause a subsequent break in the pipe in the vicinity of a plastic hinge even when the wall is weakened by a 60 circumferential crack of a depth of 30% of the wall thickness on the tension side. This pipe behavior is the result of plastic buckling in the compression side and applies to pipes of diameter-to-thickness ratio larger than 20. For this type of pipe, the axial strains decrease with increasing diameter-to-thickness ratio in the tension side. As the pipe is only loaded in one direction, there is no cyclic behavior that can trigger a subsequent break. (orig.)

On the failure probability of the primary piping of the PWR

International Nuclear Information System (INIS)

Schueller, G.I.; Hampl, N.C.

1984-01-01

A methodology for quantification of the structural reliability of the primary piping (PP) of a PWR under operational and accidental conditions is developed. Biblis B is utilized as reference plant. The PP structure is modeled utilizing finite element procedures. Based on the properties of the operational and internal accidental conditions, a static analysis suffices. However, a dynamic analysis considering non-linear effects of the soil-structure-interaction is to be used to determine load effects due to earthquake induced loading. Considering realistically the presence of initial cracks in welds and considering annual frequencies of occurrence of the various loading conditions, a crack propagation calculation utilizing the Forman model is carried out. Simultaneously leak and break probabilities using the 'Two Criteria'-Aproach are computed. A Monte Carlo simulation procedure is used as method of solution. (Author) [pt

Specialist meeting on leak before break in reactor piping and vessels

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

Study (Prediction of Main Pipes Break Rates in Water Distribution Systems Using Intelligent and Regression Methods

Directory of Open Access Journals (Sweden)

Massoud Tabesh

2011-07-01

Full Text Available Optimum operation of water distribution networks is one of the priorities of sustainable development of water resources, considering the issues of increasing efficiency and decreasing the water losses. One of the key subjects in optimum operational management of water distribution systems is preparing rehabilitation and replacement schemes, prediction of pipes break rate and evaluation of their reliability. Several approaches have been presented in recent years regarding prediction of pipe failure rates which each one requires especial data sets. Deterministic models based on age and deterministic multi variables and stochastic group modeling are examples of the solutions which relate pipe break rates to parameters like age, material and diameters. In this paper besides the mentioned parameters, more factors such as pipe depth and hydraulic pressures are considered as well. Then using multi variable regression method, intelligent approaches (Artificial neural network and neuro fuzzy models and Evolutionary polynomial Regression method (EPR pipe burst rate are predicted. To evaluate the results of different approaches, a case study is carried out in a part ofMashhadwater distribution network. The results show the capability and advantages of ANN and EPR methods to predict pipe break rates, in comparison with neuro fuzzy and multi-variable regression methods.

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

International Nuclear Information System (INIS)

1989-01-01

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

Main factors for fatigue failure probability of pipes subjected to fluid thermal fluctuation

International Nuclear Information System (INIS)

Machida, Hideo; Suzuki, Masaaki; Kasahara, Naoto

2015-01-01

It is very important to grasp failure probability and failure mode appropriately to carry out risk reduction measures of nuclear power plants. To clarify the important factors for failure probability and failure mode of pipes subjected to fluid thermal fluctuation, failure probability analyses were performed by changing the values of a stress range, stress ratio, stress components and threshold of stress intensity factor range. The important factors for the failure probability are range, stress ratio (mean stress condition) and threshold of stress intensity factor range. The important factor for the failure mode is a circumferential angle range of fluid thermal fluctuation. When a large fluid thermal fluctuation acts on the entire circumferential surface of the pipe, the probability of pipe breakage increases, calling for measures to prevent such a failure and reduce the risk to the plant. When the circumferential angle subjected to fluid thermal fluctuation is small, the failure mode of piping is leakage and the corrective maintenance might be applicable from the viewpoint of risk to the plant. (author)

Vacuum Bellows, Vacuum Piping, Cryogenic Break, and Copper Joint Failure Rate Estimates for ITER Design Use

Energy Technology Data Exchange (ETDEWEB)

L. C. Cadwallader

2010-06-01

The ITER international project design teams are working to produce an engineering design in preparation for construction of the International Thermonuclear Experimental Reactor (ITER) tokamak. During the course of this work, questions have arisen in regard to safety barriers and equipment reliability as important facets of system design. The vacuum system designers have asked several questions about the reliability of vacuum bellows and vacuum piping. The vessel design team has asked about the reliability of electrical breaks and copper-copper joints used in cryogenic piping. Research into operating experiences of similar equipment has been performed to determine representative failure rates for these components. The following chapters give the research results and the findings for vacuum system bellows, power plant stainless steel piping (amended to represent vacuum system piping), cryogenic system electrical insulating breaks, and copper joints.

Large break frequency for the SRS (Savannah River Site) production reactor process water system

International Nuclear Information System (INIS)

Daugherty, W.L.; Awadalla, N.G.; Sindelar, R.L.; Bush, S.H.

1989-01-01

The objective of this paper is to present the results and conclusions of an evaluation of the large break frequency for the process water system (primary coolant system), including the piping, reactor tank, heat exchangers, expansion joints and other process water system components. This evaluation was performed to support the ongoing PRA effort and to complement deterministic analyses addressing the credibility of a double-ended guillotine break. This evaluation encompasses three specific areas: the failure probability of large process water piping directly from imposed loads, the indirect failure probability of piping caused by the seismic-induced failure of surrounding structures, and the failure of all other process water components. The first two of these areas are discussed in detail in other papers. This paper primarily addresses the failure frequency of components other than piping, and includes the other two areas as contributions to the overall process water system break frequency

Fracture mechanics assessment of thermal aged nuclear piping based on the Leak-Before-Break concept

Energy Technology Data Exchange (ETDEWEB)

Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China); Yu, Weiwei [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China); Qian, Guian [Paul Scherrer Institute, Nuclear Energy and Safety Department, Villigen PSI (Switzerland); Wang, Rongshan; Lu, Feng; Zhang, Guodong; Xue, Fei; Chen, Zhilin [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China)

2016-05-15

Highlights: • The effects of thermal aging on crack unstable tearing are studied. • The critical size of crack unstable tearing is calculated by different methods. • The critical failure models are compared. • The conservatism of J–T diagram is shown. - Abstract: The Leak-Before-Break (LBB) concept has been accepted to design the primary piping system of the pressurized water reactor (PWR). Due to thermal aging of long term operation, the cast stainless steels (CSSs) which are used for the primary piping of PWR, suffer a significant loss of fracture toughness, and as a consequence the safety margin of the thermal aged pipe decreases. Therefore, the aged piping should be analyzed and validated by the LBB concept. In this paper, elastic–plastic fracture mechanics (EPFM) assessments of the thermal aged piping are presented according to the LBB concept. The critical break size of crack unstable tearing is calculated by the EPFM method. The crack driving force diagram (J–a diagram), the stability assessment diagram (J–T diagram) and a numerical method are applied to calculate the critical crack size of crack break. The effects of thermal aging on the plastic limit load, J–T diagram, critical crack size of the EPFM and the critical failure mode are studied. The results show that the thermal aging effect decreases the maximum allowed J-integral at a certain ductile tearing modulus by more than 50% and it increases the flow stress and plastic limit load by 11.78%. The results based on the J–T diagram are about 40% conservative than those based on the direct numerical method for the high loading case. For the thermal aged piping, it is important to consider the competition failure modes between plastic collapse and unstable ductile tearing.

Leak before break evaluation for main steam piping system made of SA106 Gr.C

Energy Technology Data Exchange (ETDEWEB)

Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

1997-04-01

The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

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

Reactor materials program process water component failure probability

International Nuclear Information System (INIS)

Daugherty, W. L.

1988-01-01

The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

Probability function of breaking-limited surface elevation. [wind generated waves of ocean

Science.gov (United States)

Tung, C. C.; Huang, N. E.; Yuan, Y.; Long, S. R.

1989-01-01

The effect of wave breaking on the probability function of surface elevation is examined. The surface elevation limited by wave breaking zeta sub b(t) is first related to the original wave elevation zeta(t) and its second derivative. An approximate, second-order, nonlinear, non-Gaussian model for zeta(t) of arbitrary but moderate bandwidth is presented, and an expression for the probability density function zeta sub b(t) is derived. The results show clearly that the effect of wave breaking on the probability density function of surface elevation is to introduce a secondary hump on the positive side of the probability density function, a phenomenon also observed in wind wave tank experiments.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

International Nuclear Information System (INIS)

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

1995-04-01

Regulatory Guide 1.45, open-quotes Reactor Coolant Pressure Boundary Leakage Detection Systems,close quotes was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, open-quotes Leak Before Break Evaluation Proceduresclose quotes where a margin of 10 on the leak detection limit is used in determining the crack size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break

The application of leak before break concepts piping of KWU-plants

International Nuclear Information System (INIS)

Bartholome, G.; Bieselt, R.W.

1985-01-01

The fracture of pipes with longitudinal and circumferential cracks was investigated by experiments and theoretical approaches (flow stress criteria and limit load analyses). The experiments show that the critical crack dimensions can conservatively be determined by fracture mechanics. The tests and calculations are applied to KWU primary coolant piping with hypothetical longitudinal and circumferential defects. Reactor systems, design, fabrication, stress analysis, material, non-destructive testing, quality control and inservice inspection are considered referring to the leak-before-break behaviour. On the basis of the extreme toughness of the materials, the known loads, the high level of non-destructive examinations, the leakage monitoring system and the high quality of manufacture and processing it is shown that a spontaneous failure need not be postulated. (orig.)

Reliability assessment for thickness measurements of pipe wall using probability of detection

International Nuclear Information System (INIS)

Nakamoto, Hiroyuki; Kojima, Fumio; Kato, Sho

2013-01-01

This paper proposes a reliability assessment method for thickness measurements of pipe wall using probability of detection (POD). Thicknesses of pipes are measured by qualified inspectors with ultrasonic thickness gauges. The inspection results are affected by human factors of the inspectors and include some errors, because the inspectors have different experiences and frequency of inspections. In order to ensure reliability for inspection results, first, POD evaluates experimental results of pipe-wall thickness inspection. We verify that the results have differences depending on inspectors including qualified inspectors. Second, two human factors that affect POD are indicated. Finally, it is confirmed that POD can identify the human factors and ensure reliability for pipe-wall thickness inspections. (author)

Leak-before-break analysis of thermally aged nuclear pipe under different bending moments

Energy Technology Data Exchange (ETDEWEB)

Lv, Xuming; Li, Shilei; Zhang, Hailong; Wang, Yanli; Wang, Xitao [University of Science and Technology Beijing, Beijing (China); Wang, Zhaoxi [CPI Nuclear Power Institute, Beijing (China); Xue, Fei [Suzhou Nuclear Power Research Institute, Suzhou (China)

2015-10-15

Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from 280°C to 450°C. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elastic–plastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Determination of times maximum insulation in case of internal flooding by pipe break

International Nuclear Information System (INIS)

Varas, M. I.; Orteu, E.; Laserna, J. A.

2014-01-01

This paper demonstrates the process followed in the preparation of the Manual of floods of Cofrentes NPP to identify the allowed maximum time available to the central in the isolation of a moderate or high energy pipe break, until it affects security (1E) participating in the safe stop of Reactor or in pools of spent fuel cooling-related equipment , and to determine the recommended isolation mode from the point of view of the location of the break or rupture, of the location of the 1E equipment and human factors. (Author)

Leak-before-break due to fatigue cracks in the cold leg piping system

International Nuclear Information System (INIS)

Mayfield, M.E.; Collier, R.P.

1984-01-01

This review paper presents the results of a deterministic assessment of the margin of safety against a large break in the cold leg piping system of pressurized water reactors. The paper focuses on the computation of leak rates resulting from fatigue cracks that penetrate the full wall thickness. Results are presented that illustrate the sensitivity of the leak rate to stress level, crack shape and crack orientation. Further, the leak rates for specific conditions are contrasted to detection levels, shutdown criteria, make-up capacity and the leak rate associated with final failure of the piping system. The results of these computations indicate that, in general, leaks far in excess of the present detection sensitivities would result at crack sizes well below the critical crack sizes for the upset loadings on the cold leg piping system

Aspects of leak before break quantification in pressurized pipes

International Nuclear Information System (INIS)

Hellen, R.A.J.; Darlaston, B.J.L.; Connors, D.C.

1980-01-01

In fitness for purpose studies of pressurized structures containing defects, the concept of leak before break is often invoked. As the assumptions used in the concept are sometimes very pessimistic it is desirable to be able to quantify them more precisely. Two aspects are currently receiving attention; these are the way in which a crack profile develops during fatigue and what happens when the remaining ligament below the crack fails. These aspects are being evaluated experimentally and theoretically. Data are presented from tests on pipes subjected to cyclic pressure and subsequently failed. An analytical approach is proposed on the question of ligament failure, this being based on the development of some recent work on flat plates. The overall question of leak before break is considered. As the understanding and confidence increases, it is possible to reduce the range of interest and focus on specific aspects of the problem. This paper examines these aspects. (author)

A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.

A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

International Nuclear Information System (INIS)

Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S.

1997-01-01

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis

Use of probabilistic methods for estimating failure probabilities and directing ISI-efforts

Energy Technology Data Exchange (ETDEWEB)

Nilsson, F; Brickstad, B [University of Uppsala, (Switzerland)

1988-12-31

Some general aspects of the role of Non Destructive Testing (NDT) efforts on the resulting probability of core damage is discussed. A simple model for the estimation of the pipe break probability due to IGSCC is discussed. It is partly based on analytical procedures, partly on service experience from the Swedish BWR program. Estimates of the break probabilities indicate that further studies are urgently needed. It is found that the uncertainties about the initial crack configuration are large contributors to the total uncertainty. Some effects of the inservice inspection are studied and it is found that the detection probabilities influence the failure probabilities. (authors).

Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

International Nuclear Information System (INIS)

Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

2014-01-01

This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

A simplified leak-before-break evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A.; Ghassemi, B. [NOVETECH Corp., Rockville, MD (United States)

1994-10-01

A simplified procedure has been defined for computing the allowable circumferential throughwall crack length as a function of applied loads in piping. This procedure has been defined to enable leak-before-break (LBB) evaluations to be performed without complex and time consuming analyses. The development of the LBB evaluation procedure is similar to that now used in Section 11 of the ASME Code for evaluation of part-throughwall flaws found in piping. The LBB evaluation procedure was bench marked using experimental data obtained from pipes having circumferential throughwall flaws. Comparisons of the experimental and predicted load carrying capacities indicate that the method has a conservative bias, such that for at least 97% of the experiments the experimental load is equal to or greater than 90% of the predicted load. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austenitic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

Some research in the field of leak before break criteria for piping

International Nuclear Information System (INIS)

Lazzeri, L.

1984-01-01

Leak-before-break research activity has lead to the following basic results: a) From an extensive analysis of the available experimental data it is concluded that the concept of net section collapse is a simple, reliable, valid tool in the case of very ductile materials. b) The analysis of some experimental data has lead to the conclusion that for partially ductile materials mixed ductile, fragile conditions may be present. c) From the analyses at a and b criteria have been established in order to compute collapse conditions for through cracked pipes as a function of the applied load (moment and axial load) in terms of net section collapse. d) The role of the thermal and secondary self equilibrating loads is discussed. e) The leak areas are often evaluated on the basis of the 0.1 Aflow criterion, i.e. somewhat arbitrarily assuming a leak area equal to 10% the pipe flow area. f) The 0.1 Aflow criteria is applied to typical lines, and it is concluded that such loads can be taken without using the classical pipe whip restraints, even if some increase in the size of the snubbers might be necessary

Modal analysis of main steam line piping under high energy line break condition

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae-Jin; Kim, Seung Hyun; Je, Sang-Yun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

If HELB (High Energy Line Break) occurs in NPPs (Nuclear Power Plants), not only environmental effect like release of radioactive material but also secondary structural defects should be considered. Jet impingement phenomenon caused by sudden pipe rupture may lead to severe damage on neighboring safe-related components and other structure. Lots of studies have been conducted to assess dynamic behaviors of the SG and MSL piping while pipe whip restraints and jet impingement shields are taken into account during design stage. Arroyo et al. performed modal analyses of a simple square component to examine the jet impingement phenomenon. Also, structural characteristics were predicted to assure structural integrity against the HELB. In this study, we examined dynamic characteristics of SG and MSL piping in a typical 1000MWe NPP. Simulation was performed by using two commercial computational softwares. In particular, modal analyses were conducted to determine mode shapes and natural frequencies of the structure and maximum displacements. The data obtain from each software were compared and observation was discussed in relation to the jet impingement phenomenon. In this research, modal analyses on the SG and MSL piping were carried out to get natural frequencies, vibration mode shapes and maximum displacements. Thereby, the following key finding was observed. (1) Maximum displacement was calculated at the top of SG outlet nozzle with y-directional bending at the third mode. (2) The differences between two models were respectively 7% in natural frequencies and less than 1% in maximum displacements.

Leak-before-break diagrams using simple plastic limit load criteria for pipes with circumferential cracks

International Nuclear Information System (INIS)

Goerner, F.; Munz, D.

1984-01-01

Simple criteria for local and global instabilities were used to calculate leak-before-break-diagrams for load-controlled deformations. Relations between the tension and bending stresses in the uncracked pipe and the critical crack angle α/sub c/, below which complete fracture cannot occur, were developed for combined loading by internal pressure and external tension and bending. The different assumptions made for local and global instability lead to similar conclusions about the allowable crack length for leak-before-break behavior. It was not the intention of this paper to compare the conclusions with experimental results available

Development on methods for evaluating structure reliability of piping components

International Nuclear Information System (INIS)

Schimpfke, T.; Grebner, H.; Peschke, J.; Sievers, J.

2003-01-01

In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour, GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The development is based on the experience achieved with applications of the public available US code PRAISE 3.10 (Piping Reliability Analysis Including Seismic Events), which was supplemented by additional features regarding the statistical evaluation and the crack orientation. PROST is designed to be more flexible to changes and supplementations. Up to now it can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents a parametric study on the influence by changing the method of stress intensity factor and limit load calculation and the statistical evaluation options on the leak probability of an exemplary pipe with postulated axial crack distribution. Furthermore the resulting leak probability of an exemplary pipe with postulated circumferential crack distribution is compared with the results of the modified PRAISE computer program. The intention of this investigation is to show trends. Therefore the resulting absolute values for probabilities should not be considered as realistic evaluations. (author)

Application of the leak-before-break concept to the primary circuit piping of the Leningrad NPP

Energy Technology Data Exchange (ETDEWEB)

Eperin, A.P.; Zakharzhevsky, Yu.O.; Arzhaev, A.I. [and others

1997-04-01

A two-year Finnish-Russian cooperation program has been initiated in 1995 to demonstrate the applicability of the leak-before-break concept (LBB) to the primary circuit piping of the Leningrad NPP. The program includes J-R curve testing of authentic pipe materials at full operating temperature, screening and computational LBB analyses complying with the USNRC Standard Review Plan 3.6.3, and exchange of LBB-related information with emphasis on NDE. Domestic computer codes are mainly used, and all tests and analyses are independently carried out by each party. The results are believed to apply generally to RBMK type plants of the first generation.

Effective applied moment in circumferential through-wall cracked pipes for leak-before-break evaluation considering pipe restraint effects

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeji; Hwang, Il-Soon [Seoul National University, Seoul 08826 (Korea, Republic of); Oh, Young-Jin, E-mail: yjoh2@kepco-enc.com [KEPCO Engineering and Construction Co. Inc., Gimcheon 39660 (Korea, Republic of)

2016-05-15

Highlights: • Effective applied moment at pipe cracked section considering the pipe restraint effect. • Verification of the proposed evaluation methods using finite element analyses. • Applicability for distributed external load of the proposed methods. - Abstract: In the leak-before-break (LBB) design of nuclear power plants, crack opening displacement (COD) is an essential element for determining the length of the leakage size crack. Recent researches regarding the evaluation of COD have indicated that the current practice of the LBB evaluation without consideration of the pressure induced bending (PIB) restraint overestimates COD, which in turn gives non-conservative results. Under a free-ended boundary condition, however, the applied moment at cracked section also can be overestimated, which has conservative effects on LBB evaluation. Therefore, it is necessary to evaluate pipe restraint effects on the applied moment as well as on COD to keep the constancy. In this paper, an evaluation method for the effect of the PIB restraint on COD and an effective applied moment (=crack driving force) at cracked section was developed. Both the linear elastic and elastic–plastic behaviors of the crack were considered. By comparing the behaviors with 3-D finite element analysis results from earlier studies, it was confirmed that the proposed methods make accurate estimations of the PIB restraint effect on COD. Next, the applicability of the proposed method to other types of external loading conditions was examined.

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)

Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

Energy Technology Data Exchange (ETDEWEB)

Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

2005-08-15

This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

Research program plan: piping. Volume 3

International Nuclear Information System (INIS)

Vagins, M.; Strosnider, J.

1985-07-01

Regulatory issues related to piping can be divided into the three areas of pipe cracking, postulated design basis pipe breaks, and design of piping for seismic and other dynamic loads. The first two of these issues are in the domain of the Materials Engineering Branch (MEBR), while the last of the three issues is the responsibility of the Mechanical/Structural Engineering Branch. This volume of the MEBR Research Plan defines the critical aspects of the pipe cracking and postulated design basis pipe break issues and identifies those research efforts and results necessary for their resolution. In general, the objectives of the MERB Piping Research Program are to provide experimentally validated analytic techniques and appropriate material properties characterization methods and data to support regulatory activities related to evaluating and ensuring piping integrity

ProLBB - A Probabilistic Approach to Leak Before Break Demonstration

Energy Technology Data Exchange (ETDEWEB)

Dillstroem, Peter; Weilin Zang (Inspecta Technology AB, Stockholm (SE))

2007-11-15

Recently, the Swedish Nuclear Power Inspectorate has developed guidelines on how to demonstrate the existence of Leak Before Break (LBB). The guidelines, mainly based on NUREG/CR-6765, define the steps that must be fulfilled to get a conservative assessment of LBB acceptability. In this report, a probabilistic LBB approach is defined and implemented into the software ProLBB. The main conclusions, from the study presented in this report, are summarized below. - The probabilistic approach developed in this study was applied to different piping systems in both Boiler Water Reactors (BWR) and Pressurised Water Reactors (PWR). Pipe sizes were selected so that small, medium and large pipes were included in the analysis. The present study shows that the conditional probability of fracture is in general small for the larger diameter pipes when evaluated as function of leak flow rate. However, when evaluated as function of fraction of crack length around the circumference, then the larger diameter pipes will belong to the ones with the highest conditional fracture probabilities. - The total failure probability, corresponding to the product between the leak probability and the conditional fracture probability, will be very small for all pipe geometries when evaluated as function of fraction of crack length around the circumference. This is mainly due to a small leak probability which is consistent with expectations since no active damage mechanism has been assumed. - One of the objectives of the approach was to be able to check the influence of off-centre cracks (i.e. the possibility that cracks occur randomly around the pipe circumference). To satisfy this objective, new stress intensity factor solutions for off-centre cracks were developed. Also to check how off-centre cracks influence crack opening areas, new form factors solutions for COA were developed taking plastic deformation into account. - The influence from an off-center crack position on the conditional

ProLBB - A Probabilistic Approach to Leak Before Break Demonstration

International Nuclear Information System (INIS)

Dillstroem, Peter; Weilin Zang

2007-11-01

Recently, the Swedish Nuclear Power Inspectorate has developed guidelines on how to demonstrate the existence of Leak Before Break (LBB). The guidelines, mainly based on NUREG/CR-6765, define the steps that must be fulfilled to get a conservative assessment of LBB acceptability. In this report, a probabilistic LBB approach is defined and implemented into the software ProLBB. The main conclusions, from the study presented in this report, are summarized below. - The probabilistic approach developed in this study was applied to different piping systems in both Boiler Water Reactors (BWR) and Pressurised Water Reactors (PWR). Pipe sizes were selected so that small, medium and large pipes were included in the analysis. The present study shows that the conditional probability of fracture is in general small for the larger diameter pipes when evaluated as function of leak flow rate. However, when evaluated as function of fraction of crack length around the circumference, then the larger diameter pipes will belong to the ones with the highest conditional fracture probabilities. - The total failure probability, corresponding to the product between the leak probability and the conditional fracture probability, will be very small for all pipe geometries when evaluated as function of fraction of crack length around the circumference. This is mainly due to a small leak probability which is consistent with expectations since no active damage mechanism has been assumed. - One of the objectives of the approach was to be able to check the influence of off-centre cracks (i.e. the possibility that cracks occur randomly around the pipe circumference). To satisfy this objective, new stress intensity factor solutions for off-centre cracks were developed. Also to check how off-centre cracks influence crack opening areas, new form factors solutions for COA were developed taking plastic deformation into account. - The influence from an off-center crack position on the conditional

Pipe-to-pipe impact program

International Nuclear Information System (INIS)

Alzheimer, J.M.; Bampton, M.C.C.; Friley, J.R.; Simonen, F.A.

1984-06-01

This report documents the tests and analyses performed as part of the Pipe-to-Pipe Impact (PTPI) Program at the Pacific Northwest Laboratory. This work was performed to assist the NRC in making licensing decisions regarding pipe-to-pipe impact events following postulated breaks in high energy fluid system piping. The report scope encompasses work conducted from the program's start through the completion of the initial hot oil tests. The test equipment, procedures, and results are described, as are analytic studies of failure potential and data correlation. Because the PTPI Program is only partially completed, the total significance of the current test results cannot yet be accurately assessed. Therefore, although trends in the data are discussed, final conclusions and recommendations will be possible only after the completion of the program, which is scheduled to end in FY 1984

Early response of pressurized hot water in a pipe to a sudden break. Final report

International Nuclear Information System (INIS)

Alamgir, M.; Kan, C.Y.; Lienhard, J.H.

1981-06-01

Experimental and analytic studies that explain the details of early pressure variations during rapid depressurization in water-cooled reactors are presented as a means of assessing sudden break consequences in a coolant pipe. The report includes (1) a description of the experiment, (2) an analysis of the new bubble growth law for thermally controlled growth of vapor bubbles in an exponentially-varying pressure field, and (3) a review of previous studies and additional observations of blowdown behavior

Follow-up Study of ITER Safety Analysis : Large In-vessel First Wall Pipe Break with Wet Confinement Bypass

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

Previous researches have been analyzed risk assessments of fusion reactors that are dangerous in the severe accidents where the radioactive material released from confinement building to the environment. To simulate the severe accidents in ITER, a number of thermal hydraulics simulation codes were used. Before construction of the fusion reactor, to obtain ITER license about safety issue, MELCOR is chosen as one of the several codes to be used to perform ITER safety analyses. Qualification of the simulation code is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. In the nuclear fusion system, the amount of released radioactive material is criteria for safety permission. Tritium (or tritiated water: HTO) and radioactive dust aerosol are the source of radioactive leakage. In the Generic Site Safety Report (GSSR) for the ITER plant, Table I lists the release guidelines for tritium and activation products for normal operation, incidents and accidents. Several accident analyses have been studied to know how much radioactive material could be released from the severe accidents. In the present work, The MELCOR input deck of large First Wall (FW) coolant leak (pipe break) is used to study and radioactive material leakage thorough bypass accident are studied to follow up the ITER safety analysis. In this research, follow-up study of the in-vessel inboard/inboard-outboard FW pipe break was analyzed to investigate the amount of leakage of radioactive aerosol. All of the accident cases released the lower amount of radioactive aerosol compared to the IAEA guide lines. In addition, the OBB pipe break made lower HTO aerosol leakage because of condensation of HTO and adsorption between coolant and aerosol.

Quantum-correlation breaking channels, quantum conditional probability and Perron-Frobenius theory

Science.gov (United States)

Chruściński, Dariusz

2013-03-01

Using the quantum analog of conditional probability and classical Bayes theorem we discuss some aspects of particular entanglement breaking channels: quantum-classical and classical-classical channels. Applying the quantum analog of Perron-Frobenius theorem we generalize the recent result of Korbicz et al. (2012) [8] on full and spectrum broadcasting from quantum-classical channels to arbitrary quantum channels.

Quantum-correlation breaking channels, quantum conditional probability and Perron–Frobenius theory

International Nuclear Information System (INIS)

Chruściński, Dariusz

2013-01-01

Using the quantum analog of conditional probability and classical Bayes theorem we discuss some aspects of particular entanglement breaking channels: quantum–classical and classical–classical channels. Applying the quantum analog of Perron–Frobenius theorem we generalize the recent result of Korbicz et al. (2012) [8] on full and spectrum broadcasting from quantum–classical channels to arbitrary quantum channels.

Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

Energy Technology Data Exchange (ETDEWEB)

Lu, S; Streit, R D; Chou, C K

1980-01-01

This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10{sup -12}). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

International Nuclear Information System (INIS)

Lu, S.; Streit, R.D.; Chou, C.K.

1980-01-01

This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

Comparison of safety margins for leak-before-break assessment of 500 MWe PHWR straight pipes: using contemporary techniques

International Nuclear Information System (INIS)

Rastogi, Rohit; Bhasin, Vivek; Kushwaha, H.S.

1998-01-01

The Leak Before Break (LBB) analysis of Primary Heat Transport (PHT) Piping of 500 MWe Indian PHWR is being performed using different well established techniques like R6 method (Nuclear Electric UK) and J-Tearing based methods (USNRC). These methods show that PHT piping has required safety margins and can be qualified for LBB. These analysis also showed that the piping has high fracture toughness and plastic collapse is the dominant mode of failure. To enhance the confidence in the results obtained from the above methods, further studies were done on the PHT piping. Procedures which predicted margins against plastic collapse were used. The analysis procedures used were Modified Limit Load Method, MPA Method (both from Germany), Moments Method (from Italy) and the Z-Factor method given in ASME Boiler and Pressure Vessel Code. The safety margins obtained from these analysis satisfied the LBB requirements. A table was generated which compared the safety margins obtained using all the above mentioned procedures. This report presents the results of this study. (author)

Reliability analysis of pipe whip impacts

International Nuclear Information System (INIS)

Alzbutas, R.; Dundulis, G.; Kulak, R.F.; Marchertas, P.V.

2003-01-01

A probabilistic analysis of a group distribution header (GDH) guillotine break and the damage resulting from the failed GDH impacting against a neighbouring wall was carried out for the Ignalita RBMK-1500 reactor. The NEPTUNE software system was used for the deterministic transient analysis of a GDH guillotine break. Many deterministic analyses were performed using different values of the random variables that were specified by ProFES software. All the deterministic results were transferred to the ProFES system, which then performed probabilistic analyses of piping failure and wall damage. The Monte Carlo Simulation (MCS) method was used to study the sensitivity of the response variables and the effect of uncertainties of material properties and geometry parameters to the probability of limit states. The First Order Reliability Method (FORM) was used to study the probability of failure of the impacted-wall and the support-wall. The Response Surface (RS/MCS) method was used in order to express failure probability as function and to investigate the dependence between impact load and failure probability. The results of the probability analyses for a whipping GDH impacting onto an adjacent wall show that: (i) there is a 0.982 probability that after a GDH guillotine break contact between GDH and wall will occur; (ii) there is a probability of 0.013 that the ultimate tensile strength of concrete at the impact location will be reached, and a through-crack may open; (iii) there is a probability of 0.0126 that the ultimate compressive strength of concrete at the GDH support location will be reached, and the concrete may fail; (iv) at the impact location in the adjacent wall, there is a probability of 0.327 that the ultimate tensile strength of the rebars in the first layer will be reached and the rebars will fail; (v) at the GDH support location, there is a probability of 0.11 that the ultimate stress of the rebars in the first layer will be reached and the rebars will fail

The detection of leaks on sodium pipes in a 'leak before break' approach

International Nuclear Information System (INIS)

Antonakas, D.

1989-01-01

The operation of circuits containing liquid sodium requires, given the chemical affinity of this fluid for air and water, a reliable detection of possible leaks. This system of detection should alert the operators to the occurrence of a leak in sufficient time to limit the potential consequences of a discharge of sodium in the building, leading to a severe sodium fire or at least to an extended corrosion of the pipe system. From a design point of view, the most likely event leading to this situation can be the consequence. of an initial undetected defect which develops under the effect of thermo-mechanical loadings, produces a sodium. leak below the dejection threshold remains undetectable white progressing and finally leads to a guillotine-type rupture when an incidental loading is superimposed to the normal one. The 'leak before break' approach which is now currently introduced in design considerations consists of insuring the detection of incipient leaks corresponding to through-the-wall cracks well below instability of the pipe. Under this short statement, lies a considerable and still necessary effort of research broadly presented in the present paper

Study on unstable fracture characteristics of light water reactor piping

International Nuclear Information System (INIS)

Kurihara, Ryoichi

1998-08-01

Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs

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.

Determination of Secondary Encasement Pipe Design Pressure

Energy Technology Data Exchange (ETDEWEB)

TEDESCHI, A.R.

2000-10-26

This document published results of iterative calculations for maximum tank farm transfer secondary pipe (encasement) pressure upon failure of the primary pipe. The maximum pressure was calculated from a primary pipe guillotine break. Results show encasement pipeline design or testing pressures can be significantly lower than primary pipe pressure criteria.

Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

International Nuclear Information System (INIS)

Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

2010-01-01

This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

Analysis of breaks in pipe connections to the hot leg and to the loop seal in the primary system of Ringhals 2 PWR

International Nuclear Information System (INIS)

Nilsson, L.; Sjoeberg, A.

1987-01-01

Analysis has been made of seven different cases of breaks in pipes connected to the hot leg and to the loop seal in Ringhals 2 PWR. The pipes, in which guillotine breaks are postulated, have nominal diameters ranging from 1 to 14 inches. The purpose of the analysis is to supplement the basis for a review of the inspection procedures for the safety of pressure vessels prescribed by SKI. A similar analysis already exists concerning breaks in the cold leg connections. The analysis has been made using the thermal hydraulic computer code RELAPS/MOD2 and with best estimate assumptions. This means that normal operating conditions have been adopted for the input to the calculations. However, the capacity of the safety injection system was assumed to be reduced by having one pump not operating each of the low pressure and high pressure safety injection system. The results of the analysis are presented in tables and as computer plots. The analysis shows that the consequences with respect to increased fuel rod and cladding temperatures are quite harmless. Only the two cases with the largest break sizes lead to critical heat flux (CHF) and increased temperatures for the hottest rods in the core. The peak cladding temperature was 636 degrees C for the 12 inch break. In both cases rewetting occurred within 25 s of accident initiation. In the cases with breaks in connections of 6 inch diameter and smaller the RELAP5 calculations indicated a substantial margin to CHF throughout the transient. (authors)

Small break critical discharge: The roles of vapor and liquid entrainment in a stratified two-phase region upstream of the break

International Nuclear Information System (INIS)

Schrock, V.E.; Revankar, S.T.; Mannheimer, R.; Wang, C.H.

1986-12-01

The main objective of this research program was to perform an experimental investigation on the phenomena of two-phase critical flow through small break from a horizontal pipe which contained a stratified two phase flow. Stagnation conditions investigated were saturated steam-water, and air-cold water at pressures ranging from 0.37 MPa to 1.07 MPa. The small breaks employed were cylindrical tubes of diameters 3.96 mm, 6.32 mm, and 10.1 mm with sharp-edged entrance. For breaks resulting from a small hole in a primary coolant pipe or in a small pipe, a sharp-edged orifice or a sharp-edged tube can be the approximation

Analysis of leak and break behavior in a failure assessment diagram for carbon steel pipes

International Nuclear Information System (INIS)

Kanno, Satoshi; Hasegawa, Kunio; Shimizu, Tasuku; Saitoh, Takashi; Gotoh, Nobuho

1992-01-01

The leak and break behavior of a cracked coolant pipe subjected to an internal pressure and a bending moment was analyzed with a failure assessment diagram using the R6 approach. This paper examines the conditions of the detectable coolant leakage without breakage. A leakage assessment curve, a locus of assessment point for detectable coolant leakage, was defined in the failure assessment diagram. The region between the leak assessment and failure assessment curves satisfies the condition of detectable leakage without breakage. In this region, a crack can be safely inspected by a coolant leak detector. (orig.)

Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

Energy Technology Data Exchange (ETDEWEB)

Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

2002-12-01

Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

Analysis methods for structure reliability of piping components

International Nuclear Information System (INIS)

Schimpfke, T.; Grebner, H.; Sievers, J.

2004-01-01

In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour (BMWA) GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The long-term objective of this development is to provide failure probabilities of passive components for probabilistic safety analysis of nuclear power plants. Up to now the code can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents some of the results of a benchmark analysis in the frame of the European project NURBIM (Nuclear Risk Based Inspection Methodologies for Passive Components). (orig.)

These Pipes Are "Happening"

Science.gov (United States)

Skophammer, Karen

2010-01-01

The author is blessed with having the water pipes for the school system in her office. In this article, the author describes how the breaking of the pipes had led to a very worthwhile art experience for her students. They practiced contour and shaded drawing techniques, reviewed patterns and color theory, and used their reasoning skills--all while…

Pipe rupture test results; 6 in. pipe whip test under BWR LOCA conditions

International Nuclear Information System (INIS)

Kurihara, Ryoichi; Yano, Toshikazu; Ueda, Shuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kato, Rokuro; Miyazono, Shohachiro

1983-02-01

A series of pipe rupture tests has been performed in JAERI to demonstrate the safety of the primary coolant circuits in the event of pipe rupture, in nuclear power plants. The present report summarizes the results of 6 in. pipe whip tests (RUN 5605, 5606), under BWR LOCA conditions (285 0 C, 6.8 MPa), which were performed in August, 1981. The test pipe is made of Type 304 stainless steel and its outer diameter is 6 in. and its thickness is 11.1 mm. The restraints are made of Type 304 stainless steel and its diameter is 16.0 mm. Two restraints were set on the restraint support with clearance of 100 mm. Overhang length was varied as the parameter in these tests and was 300 mm or 700 mm. The following results are obtained. (1) The deformations of a pipe and restraints are limited effectively by shorter overhang length of 300. However, they become larger when the overhang length is 700 mm, and the pipe deforms especially at the setting point of restraints. (2) Velocity at the free end of pipe becomes about 30 m/sec just after the break. However, velocity at the setting point of restraint becomes about only 4 m/sec just after the break. (3) It seems from the comparison between the 4 in. tests and 6 in. tests that the maximum restraint force of 6 in. tests is about two times as large as that of 4 in. tests. (author)

Experimental studies of PWR primary piping under loca

International Nuclear Information System (INIS)

Caumette, Pierre; Garcia, J.L.

1980-07-01

The experimental program performed on AQUITAINE II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It has been developed in the form of a quadripartite agreement between the Commissariat a l'Energie Atomique, Framatome, Electricite de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S- or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: - the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; - the impact function of a pipe or an energy-absorbing bumper; - the lateral stability of both ends of a pipe, after a double-guillotine break [fr

Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

Energy Technology Data Exchange (ETDEWEB)

Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

1997-04-01

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-02-15

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

Dynamic experiments on cracked pipes

International Nuclear Information System (INIS)

Petit, M.; Brunet, G.; Buland, P.

1991-01-01

In order to apply the leak before break concept to piping systems, the behavior of cracked pipes under dynamic, and especially seismic loading must be studied. In a first phase, an experimental program on cracked stainless steel pipes under quasi-static monotonic loading has been conducted. In this paper, the dynamic tests on the same pipe geometry are described. These tests have been performed on a shaking table with a mono frequency input signal. The main parameter of the tests is the frequency of excitation versus the frequency of the system

Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

International Nuclear Information System (INIS)

Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

1988-12-01

A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

Failure Analysis Of Industrial Boiler Pipe

International Nuclear Information System (INIS)

Natsir, Muhammad; Soedardjo, B.; Arhatari, Dewi; Andryansyah; Haryanto, Mudi; Triyadi, Ari

2000-01-01

Failure analysis of industrial boiler pipe has been done. The tested pipe material is carbon steel SA 178 Grade A refer to specification data which taken from Fertilizer Company. Steps in analysis were ; collection of background operation and material specification, visual inspection, dye penetrant test, radiography test, chemical composition test, hardness test, metallography test. From the test and analysis result, it is shown that the pipe failure caused by erosion and welding was shown porosity and incomplete penetration. The main cause of failure pipe is erosion due to cavitation, which decreases the pipe thickness. Break in pipe thickness can be done due to decreasing in pipe thickness. To anticipate this problem, the ppe will be replaced with new pipe

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

International Nuclear Information System (INIS)

Gamble, R.M.; Wichman, K.R.

1997-01-01

The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Wichman, K.R.

1997-04-01

The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

International Nuclear Information System (INIS)

Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

2002-01-01

The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, A cceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors . These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the

Pipe-to-pipe impact tests

Energy Technology Data Exchange (ETDEWEB)

Bampton, M C.C.; Alzheimer, J M; Friley, J R; Simonen, F A

1985-11-01

Existing licensing criteria express what damage shall be assumed for various pipe sizes as a consequence of a postulated break in a high energy system. The criteria are contained in Section 3.6.2 of the Standard Review Plan, and the purpose of the program described with this paper is to evaluate the impact criteria by means of a combined experimental and analytical approach. A series of tests has been completed. Evaluation of the test showed a deficiency in the range of test parameters. These deficiencies are being remedied by a second series of tests and a more powerful impact machine. A parallel analysis capability has been developed. This capability has been used to predict the damage for the first test series. The quality of predictions has been improved by tests that establish post-crush and bending relationships. Two outputs are expected from this project: data that may, or may not, necessitate changes to the criteria after appropriate value impact evaluations and an analytic capability for rapidly evaluating the potential for pipe whip damage after a postulated break. These outputs are to be contained in a value-impact document and a program final report. (orig.).

Leak before break behaviour of austenitic and ferritic pipes containing circumferential defects

Energy Technology Data Exchange (ETDEWEB)

Stadtmueller, W.; Sturm, D.

1997-04-01

Several research projects carried out at MPA Stuttgart to investigate the Leak-before-Break (LBB) behavior of safety relevant pressure bearing components are summarized. Results presented relate to pipes containing circumferential defects subjected to internal pressure and external bending loading. An overview of the experimentally determined results for ferritic components is presented. For components containing postulated or actual defects, the dependence of the critical loading limit on the defect size is shown in the form of LBB curves. These are determined experimentally and/or by calculation for through-wall slits, and represent the boundary curve between leakage and massive fracture. For surface defects and a given bending moment and internal pressure, no fracture will occur if the length at leakage remains smaller than the critical defect length given by the LBB curve for through-wall defects. The predictive capability of engineering calculational methods are presented by way of example. The investigation programs currently underway, testing techniques, and initial results are outlined.

Application of leak-before-break criteria to pressurized water reactors

International Nuclear Information System (INIS)

Roege, P.; Day, B.; Beckjord, E.; Golay, M.

1986-01-01

The possibility of consequential damage to safety-related systems or components after postulated pipe breaks in Light Water Reactors has led to the installation of pipe restraints capable of withstanding the loads in such an accident. These restraints are a significant part of initial capital cost, and because of their size and location, impede plant maintenance. The Piping Review Committee of the U.S. Nuclear Regulatory Commission has concluded that, subject to fulfillment of certain criteria, the pipe restraints for pressurized water reactor main coolant piping are not necessary, because the failure mode of this piping is such that it will leak before it will break, and the leakage of reactor coolant is large enough to detect. In this study, we examine the piping systems of a 4-loop 1,150 MWe pressurized water reactor, determining the crack size that would be stable from a fracture mechanics point of view, and the range of leak rates that would ensue. We then consider the sensitivity of conventional leak detection systems, and find that pipe sizes down to 45 cm in diameter would meet the leak-before-break criteria. Improvements in the sensitivity of conventional leak detectors would extend this range down to pipe sizes down to the range of 20 - 45 cm in diameter. The development of local leak detection systems which would respond to leaks in compartments or confined areas would make it possible to apply the criteria to sizes as low as 10 - 20 cm in diameter, which appear to be the limit of the net cost savings of eliminating pipe restraints and adding additional leak detection instrumentation. Extending the leak-before-break concept into this smallest pipe range may require improved precision in crack definition, flow modeling, and leak detection. Better detection of leaks may also require use of new detection methods coupled to novel approaches to piping system design. (J.P.N.)

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)

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)

Acoustic Signal Processing for Pipe Condition Assessment (WaterRF Report 4360)

Science.gov (United States)

Unique to prestressed concrete cylinder pipe (PCCP), individual wire breaks create an excitation in the pipe wall that may vary in response to the remaining compression of the pipe core. This project was designed to improve acoustic signal processing for pipe condition assessment...

Nonlinear dynamic analysis of high energy line pipe whip

International Nuclear Information System (INIS)

Hsu, L.C.; Kuo, A.Y.; Tang, H.T.

1983-01-01

To facilitate potential cost savings in pipe whip protection design, TVA conducted a 1'' high pressure line break test to investigate the pipe whip behavior. The test results are available to EPRI as a data base for a generic study on nonlinear dynamic behavior of piping systems and pipe whip phenomena. This paper describes a nonlinear dynamic analysis of the TVA high energy line tests using ABAQUS-EPGEN code. The analysis considers the effects of large deformation and high strain rate on resisting moment and energy absorption capability of the analyzed piping system. The numerical results of impact forces, impact velocities, and reaction forces at pipe supports are compared to the TVA test data. The pipe whip impact time and forces have also been calculated per the current NRC guidelines and compared. The calculated pipe support reaction forces prior to impact have been found to be in good agreement with the TVA test data except for some peak values at the very beginning of the pipe break. These peaks are believed to be due to stress wave propagation which cannot be addressed by the ABAQUS code. Both the effects of elbow crushing and strain rate have been approximately simulated. The results are found to be important on pipe whip impact evaluation. (orig.)

A Monte Carlo calculation of the pionium break-up probability with different sets of pionium target cross sections

International Nuclear Information System (INIS)

Santamarina, C; Schumann, M; Afanasyev, L G; Heim, T

2003-01-01

Chiral perturbation theory predicts the lifetime of pionium, a hydrogen-like π + π - atom, to better than 3% precision. The goal of the DIRAC experiment at CERN is to obtain and check this value experimentally by measuring the break-up probability of pionium in a target. In order to accurately measure the lifetime one needs to know the relationship between the break-up probability and the lifetime to 1% accuracy. We have obtained this dependence by modelling the evolution of pionic atoms in the target using Monte Carlo methods. The model relies on the computation of the pionium-target-atom interaction cross sections. Three different sets of pionium-target cross sections with varying degrees of complexity were used: from the simplest first-order Born approximation involving only the electrostatic interaction to a more advanced approach, taking into account multiphoton exchanges and relativistic effects. We conclude that, in order to obtain the pionium lifetime to 1% accuracy from the break-up probability, the pionium-target cross sections must be known with the same accuracy for the low excited bound states of the pionic atom. This result has been achieved, for low Z targets, with the two most precise cross section sets. For large Z targets only the set accounting for multiphoton exchange satisfies the condition

Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

Comparative study of approaches to estimate pipe break frequencies

Energy Technology Data Exchange (ETDEWEB)

Simola, K.; Pulkkinen, U.; Talja, H.; Saarenheimo, A.; Karjalainen-Roikonen, P. [VTT Industrial Systems (Finland)

2002-12-01

The report describes the comparative study of two approaches to estimate pipe leak and rupture frequencies for piping. One method is based on a probabilistic fracture mechanistic (PFM) model while the other one is based on statistical estimation of rupture frequencies from a large database. In order to be able to compare the approaches and their results, the rupture frequencies of some selected welds have been estimated using both of these methods. This paper highlights the differences both in methods, input data, need and use of plant specific information and need of expert judgement. The study focuses on one specific degradation mechanism, namely the intergranular stress corrosion cracking (IGSCC). This is the major degradation mechanism in old stainless steel piping in BWR environment, and its growth is influenced by material properties, stresses and water chemistry. (au)

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

International Nuclear Information System (INIS)

Li Chengliang; Deng Xiaoyun; Yin Zhiying; Liu Meng

2012-01-01

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

Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

International Nuclear Information System (INIS)

Dunn, F.E.

1990-01-01

The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

Leak-Before-Break assessment of a welded piping based on 3D finite element method

International Nuclear Information System (INIS)

Chen, Mingya; Yu, Weiwei; Chen, Zhilin; Qian, Guian; Lu, Feng; Xue, Fei

2017-01-01

Highlights: • The effects of load reduction, strength match, welding width, load level, crack size and constraint are studied. • The results show that the LBB margin is dependent on the load level. • The results show that higher strength-match of WPJs will have higher crack-front constraints. • The results show that the engineering method has a high precision only if the width of weld is comparable to the crack depth. - Abstract: The paper studies the effects of the load reduction (discrepancy between designing and real loadings), strength match of the welded piping joint (WPJ), welding width, crack size and crack tip constraint on the Leak-Before-Break (LBB) assessment of a welded piping. The 3D finite element (FE) method is used in the study of a surge line of the steam generator in a nuclear power plant. It is demonstrated that the LBB margin is dependent on the loading level and the load reduction effect should be considered. When the loading is high enough, there is a quite large deviation between the J-integral calculated based on the real material property of WPJ and that calculated based on the engineering method, e.g. Zahoor handbook of Electric Power Research Institute (EPRI). The engineering method assumes that the whole piping is made of the unique welding material in the calculation. As the influence of the strength matching and welding width is ignored in the engineering method for J-integral calculation, the engineering method has a sufficient precision only if the width of welding is comparable to the crack depth. Narrower welding width leads to higher constraint of the plastic deformation in the welding and larger high stress areas in the base for the low strength-match WPJ. Higher strength matching of WPJs has higher crack-front constraints.

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.)

Evaluation of the probability of crack initiation and crack instability for a pipe with a semi-elliptical crack

International Nuclear Information System (INIS)

Le Delliou, P.; Hornet, P.

2001-01-01

This paper presents some work conducted at EDF R and D Division to evaluate the probability that a semi-elliptical crack in a pipe not only initiates but also propagates when submitted to mechanical loading such as bending and pressure combined or not with a thermal shock. The first part is related to the description of the mechanical model: the simplified methods included in the French RSE-M Code used to evaluate the J-integral as well as the principle of the determination of the crack propagation. Then, the way this deterministic approach is combined to a reliability code is described. Finally, an example is shown: the initiation and the instability of a semi-elliptical crack in a pipe submitted to combined pressure and bending moment. (author)

Device for achieving pressure balance in the steam generator of a power plant in case of a main-steam pipe or a feedwater pipe break

International Nuclear Information System (INIS)

Wietelmann, F.

1978-01-01

In order to increase the safety in the steam generator of a power plant in case of a pipe break, the possibility of a pressure balance between the feedwater inlet and the initial steam outlet chambers is allowed for. According to the invention, the partition wall separating these two chambers will exhibit several overflow openings, each of which will be provided with a closure and half of which may be opened to one side only, care having been taken that in case of an accident on occurrence of a certain differential pressure they will always be opened to the low-pressure side. As closures caps, which may be swing out of the way, or rupture diaphragms are mentioned. (UWI) 891 HP [de

Detection probability of least tern and piping plover chicks in a large river system

Science.gov (United States)

Roche, Erin A.; Shaffer, Terry L.; Anteau, Michael J.; Sherfy, Mark H.; Stucker, Jennifer H.; Wiltermuth, Mark T.; Dovichin, Colin M.

2014-01-01

Monitoring the abundance and stability of populations of conservation concern is often complicated by an inability to perfectly detect all members of the population. Mark-recapture offers a flexible framework in which one may identify factors contributing to imperfect detection, while at the same time estimating demographic parameters such as abundance or survival. We individually color-marked, recaptured, and re-sighted 1,635 federally listed interior least tern (Sternula antillarum; endangered) chicks and 1,318 piping plover (Charadrius melodus; threatened) chicks from 2006 to 2009 at 4 study areas along the Missouri River and investigated effects of observer-, subject-, and site-level covariates suspected of influencing detection. Increasing the time spent searching and crew size increased the probability of detecting both species regardless of study area and detection methods were not associated with decreased survival. However, associations between detection probability and the investigated covariates were highly variable by study area and species combinations, indicating that a universal mark-recapture design may not be appropriate.

Additional requirements for leak-before-break application to primary coolant piping in Belgium

Energy Technology Data Exchange (ETDEWEB)

Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

1997-04-01

Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

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

International Nuclear Information System (INIS)

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

1981-08-01

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

Pipe rupture test results: 4-inch pipe whip tests under PWR LOCA conditions

International Nuclear Information System (INIS)

Miyazaki, Noriyuki; Ueda, Shuzo; Isozaki, Toshikuni; Kato, Rokuro; Kurihara, Ryoichi; Yano, Toshikazu; Miyazono, Shohachiro

1982-09-01

This report summarizes the results of 4-inch pipe whip tests (RUN No. 5506, 5507, 5508 and 5604) under the PWR LOCA conditions. The dynamic behaviors of the test pipe and restraints were studied in the tests. In the tests, the gap between the test pipe and the restraints was kept at the constant value of 8.85 mm and the overhang length was varied from 250 mm to 650 mm. The dynamic behaviors of the test pipe and the restraint were made clear by the outputs of strain gages and the measurements of residual deformations. The data of water hammer in subcooled water were also obtained by the pressure transducers mounted on the test pipe. The main conclusions obtained from the tests are as follows. (1) The whipping of pipe can be prevented more effectively as the overhang length becomes shorter. (2) The load acting on the restraint-support structure becomes larger as the overhang length becomes shorter. (3) The restraint farther from the break location does not limit the pipe movement except for the first impact when the overhang length is long. (4) The ultimate moment M sub(u) of the pipe at the restraint location can be used to predict the plastic collapse of the whipping pipe. (5) The restraints slide along the pipe axis and are subjected to bending moment, when the overhang length is long. (author)

Proceedings of the seminar on leak before break in reactor piping and vessels

International Nuclear Information System (INIS)

Faidy, C.; Gilles, P.

1997-04-01

The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the seminar on leak before break in reactor piping and vessels

Energy Technology Data Exchange (ETDEWEB)

Faidy, C. [ed.] [Electricite de France, Villeurbanne (France); Gilles, P. [ed.] [Framatome, Paris (France)

1997-04-01

The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

A multi-step approach for evaluation of pipe impact effects

International Nuclear Information System (INIS)

Vazquez Sierra, J.M.; Marti, J.

1987-01-01

The licensing of new and requalification of existing plant requires the consideration of effects arising from postulated breaks in high-energy lines. If the resulting jets or whipping pipes affect equipment or components (with safety-related functions in relation with the postulated break), their structural integrity and functionality has to be guaranteed. This can be achieved either by demonstrating sufficient ruggedness, or by obviating the problem with hardware (restraints, screens, deflectors, etc.). The present paper is orientated towards the first solution. A methodology has been developed and applied to the requalification of high-energy piping at the Santa Maria de Garona NPP in Spain. It provides techniques for evaluation of pipe-whip and jet effects on various structures inside the containment: containment liner, pedestal, shield wall, pipes and penetrations. Items of little structural strength (such as cables, conduits, etc.) were excluded from this approach for obvious reasons. (orig./GL)

Computer-Aided Analysis of Flow in Water Pipe Networks after a Seismic Event

Directory of Open Access Journals (Sweden)

Won-Hee Kang

2017-01-01

Full Text Available This paper proposes a framework for a reliability-based flow analysis for a water pipe network after an earthquake. For the first part of the framework, we propose to use a modeling procedure for multiple leaks and breaks in the water pipe segments of a network that has been damaged by an earthquake. For the second part, we propose an efficient system-level probabilistic flow analysis process that integrates the matrix-based system reliability (MSR formulation and the branch-and-bound method. This process probabilistically predicts flow quantities by considering system-level damage scenarios consisting of combinations of leaks and breaks in network pipes and significantly reduces the computational cost by sequentially prioritizing the system states according to their likelihoods and by using the branch-and-bound method to select their partial sets. The proposed framework is illustrated and demonstrated by examining two example water pipe networks that have been subjected to a seismic event. These two examples consist of 11 and 20 pipe segments, respectively, and are computationally modeled considering their available topological, material, and mechanical properties. Considering different earthquake scenarios and the resulting multiple leaks and breaks in the water pipe segments, the water flows in the segments are estimated in a computationally efficient manner.

Conceptual design of pipe whip restraints using interactive computer analysis

International Nuclear Information System (INIS)

Rigamonti, G.; Dainora, J.

1975-01-01

Protection against pipe break effects necessitates a complex interaction between failure mode analysis, piping layout, and structural design. Many iterations are required to finalize structural designs and equipment arrangements. The magnitude of the pipe break loads transmitted by the pipe whip restraints to structural embedments precludes the application of conservative design margins. A simplified analytical formulation of the nonlinear dynamic problems associated with pipe whip has been developed and applied using interactive computer analysis techniques. In the dynamic analysis, the restraint and the associated portion of the piping system, are modeled using the finite element lumped mass approach to properly reflect the dynamic characteristics of the piping/restraint system. The analysis is performed as a series of piecewise linear increments. Each of these linear increments is terminated by either formation of plastic conditions or closing/opening of gaps. The stiffness matrix is modified to reflect the changed stiffness characteristics of the system and re-started using the previous boundary conditions. The formation of yield hinges are related to the plastic moment of the section and unloading paths are automatically considered. The conceptual design of the piping/restraint system is performed using interactive computer analysis. The application of the simplified analytical approach with interactive computer analysis results in an order of magnitude reduction in engineering time and computer cost. (Auth.)

Application of Leak Before Break concept in 316LN austenitic steel pipes welded using 316L

International Nuclear Information System (INIS)

Cunto, Gabriel Giannini de

2017-01-01

This work presents a study of application of the Leak Before Break (LBB) concept, usually applied in nuclear power plants, in a pipe made from steel AISI type 316LN welded a coated electrode AISI type 316L. LBB concept is a criterion based on fracture mechanics analysis to show that a crack leak, present in a pipe, can be detected by leak detection systems, before this crack reaches a critical size that results in pipe fail. In the studied pipe, tensile tests and Ramberg-Osgood analyses were performed, as well as fracture toughness tests for obtaining the material resistance curve J-R. The tests were performed considering the base metal, weld and heat affected zone (HAZ), at the same operating temperatures of a nuclear power plant. For the mechanical properties found in these tests, load limit analyses were performed in order to determine the size of a crack which could cause a detectable leakage and the critical crack size, considering failure by plastic collapse. For the critical crack size found in the weld, which is the region that presented the lowest toughness, Integral J and tearing modulus T analyses were performed, considering failure by tearing instability. Results show a well-defined behavior between the base metal, HAZ and weld zones, where the base metal has a high toughness behavior, the weld has a low toughness behavior and the HAZ showed intermediate mechanical properties between the base metal and the weld. Using the PICEP software, the leak rate curves versus crack size and also the critical crack size were determined by considering load limit analysis. It was observed that after a certain crack size, the leak rate in base metal is much higher than for the HAZ and the weld, considering the same crack length. This occurs because in the base metal crack, it is expected that the crack grows in a more rounded form due to its higher toughness. The lowest critical crack size was found for the base metal presenting circumferential cracks. For the

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

A spatial decision support system for pipe-break susceptibility ...

African Journals Online (AJOL)

lying properties. Existing decision support systems available in the field of water distribution system maintenance mainly focus on leak detection and pipe rehabilitation/replacement strategies. These existing systems, however, do not address the ...

Ductile fracture behaviour of primary heat transport piping material ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

Abstract. Design of primary heat transport (PHT) piping of pressurised heavy water reactors (PHWR) has to ensure implementation of leak-before-break con- cepts. In order to be able to do so, the ductile fracture characteristics of PHT piping material have to be quantified. In this paper, the fracture resistance of SA333, Grade.

Determination of times maximum insulation in case of internal flooding by pipe break; Determinacion de los tiempos maximos de aislamiento en caso de inundacion interna por rotura de tuberia

Energy Technology Data Exchange (ETDEWEB)

Varas, M. I.; Orteu, E.; Laserna, J. A.

2014-07-01

This paper demonstrates the process followed in the preparation of the Manual of floods of Cofrentes NPP to identify the allowed maximum time available to the central in the isolation of a moderate or high energy pipe break, until it affects security (1E) participating in the safe stop of Reactor or in pools of spent fuel cooling-related equipment , and to determine the recommended isolation mode from the point of view of the location of the break or rupture, of the location of the 1E equipment and human factors. (Author)

Development of integrated insulation joint for cooling pipe in tokamak reactor

International Nuclear Information System (INIS)

Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro.

1994-08-01

In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel/ceramics/stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured. (author)

Short cracks in piping and piping welds. Seventh program report, March 1993-December 1994. Volume 4, Number 1

Energy Technology Data Exchange (ETDEWEB)

Wilkowski, G.M.; Ghadiali, N.; Rudland, D.; Krishnaswamy, P.; Rahman, S.; Scott, P. [Battelle, Columbus, OH (United States)

1995-04-01

This is the seventh progress report of the U.S. Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks in Piping and Piping Welds{close_quotes}. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports.

Short cracks in piping and piping welds. Seventh program report, March 1993-December 1994. Volume 4, Number 1

International Nuclear Information System (INIS)

Wilkowski, G.M.; Ghadiali, N.; Rudland, D.; Krishnaswamy, P.; Rahman, S.; Scott, P.

1995-04-01

This is the seventh progress report of the U.S. Nuclear Regulatory Commission's research program entitled open-quotes Short Cracks in Piping and Piping Weldsclose quotes. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports

Rupture disc opening property for using pipe rupture test in JAERI

International Nuclear Information System (INIS)

Kato, Rokuro

1983-03-01

In the Mechanical Strength and Structure Lab of JAERI there are being performed pipe break tests which are a postulated instantaneous guillotine break of the primary coolant piping in nuclear power plants. The test being performed are pipe whip tests and jet discharging tests. The bursting of the rupture disc is initiated by an electrical arc and is concluded by the internal pressure. Because the time characteristics during the opening of the rupture disc affects the dynamic thrust force of the pipe, it is necessary to measure these time characteristics. However, it is difficult to measure the conditions during this continuous opening because at the same time of the opening the high temperature and high pressure water is flashing. Therefore, the rupture disc opening was postulated on the measuring of the effective opening characteristics with electric contraction terminals which were attached to the inner surface of the test pipe downstream of the rupture disc and were extended toward the pipe centerline in a ring whose area is about 60 % of the area of the pipe flow sectional area. The measurement voltage was recorded when the data recorder was started in sequence with the electrical arc release from a trigger signal. As a result, it is evident that under high temperature and high pressure water the effective opening time is delayed by a few milliseconds. (author)

Application of LBB to high energy piping systems in operating PWR

Energy Technology Data Exchange (ETDEWEB)

Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

1997-04-01

The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

Problems identified in quantifying leak before break in pressure containing structures

International Nuclear Information System (INIS)

Darlaston, B.J.L.; Connors, D.C.; Hellen, R.A.J.

1979-01-01

The leak before break approach is often applied to pressure containing plant as part of the safety assessment. The assumptions used in this approach are sometimes very pessimistic. It is therefore desirable to be able to quantify the concept more precisely. The two aspects which are of considerable importance are the way the crack profile develops and what happens when the remaining ligament below the crack fails. These two aspects are receiving attention and together with the development of the basic concept of 'leak before break' form the basis of this paper. Some thirty burst tests have been carried out on straight pipes of various dimensions. The results have been analysed using the CEGB Failure Assessment Route for structures containing defects. It was shown that in most cases the leaks and the breaks could be separated by this procedure. However all these tests involved machined rather than fatigue grown defects. A complementary program on pipes has the objective of examining defect growth under cyclic loads. The tests on the 152 mm diameter pipes showed that these defects did not grow in a uniform manner but after a while began to tunnel through the wall locally leading to failure of part of the ligament. This implies that some defects considered to be in the break category would only lead to leaks. As a consequence of these results the experimental programme was redesigned to concentrate on the growth of defects which it was thought would span the boundary of leak and break. For the pipe dimensions and materials used, this represented long defects which would penetrate well into the wall before ligament failure occurred. The analysis and interpretation of this aspect of the programme is part analytical part empirical. (orig.)

Reliability analysis of stainless steel piping using a single stress corrosion cracking damage parameter

International Nuclear Information System (INIS)

Guedri, A.

2013-01-01

This article presents the results of an investigation that combines standard methods of fracture mechanics, empirical correlations of stress-corrosion cracking, and probabilistic methods to provide an assessment of Intergranular Stress Corrosion Cracking (IGSCC) of stainless steel piping. This is done by simulating the cracking of stainless steel piping under IGSCC conditions using the general methodology recommended in the modified computer program Piping Reliability Analysis Including Seismic Events, and by characterizing IGSCC using a single damage parameter. Good correlation between the pipe end-life probability of leak and the damage values were found. These correlations were later used to generalize this probabilistic fracture model. Also, the probability of detection curves and the benefits of in-service inspection in order to reduce the probability of leak for nuclear piping systems subjected to IGSCC were discussed for several pipe sizes. It was found that greater benefits could be gained from inspections for the large pipe as compared to the small pipe sizes. Also, the results indicate that the use of a better inspection procedure can be more effective than a tenfold increase in the number of inspections of inferior quality. -- Highlights: • We simulate the pipe probability of failure under different level of SCC damages. • The residual stresses are adjusted to calibrate the model. • Good correlations between 40-year cumulative leak probabilities and D σ are found. • These correlations were used to generalize this probabilistic fracture model. • We assess the effect of inspection procedures and scenarios on leak probabilities

Piping hydrodynamic loads for a PWR power up-rate with steam generator replacement

International Nuclear Information System (INIS)

Julie M Jarvis; Allen T Vieira; James M Gilmer

2005-01-01

Full text of publication follows: Pipe break hydrodynamic loads are calculated for various systems in a PWR for a Power Up-rate (PUR) with a Steam Generator Replacement (SGR). PUR with SGR can change the system pressures, mass flowrates and pipe routing/configuration. These changes can alter the steam generator piping steam/water hammer loads. This paper discusses the need to benchmark against the original design basis, the use of different modeling techniques, and lessons learned. Benchmarking for licensing in the United States is vital in consideration of 10CFR50.59 and other licensing and safety issues. RELAP5 and its force post-processor R5FORCE are used to model the transient loads for various piping systems such as main feedwater and blowdown systems. Other modeling applications, including the Bechtel GAFT program, are used to evaluate loadings in the main steam piping. Forces are calculated for main steam turbine stop valve closure, feedwater pipe breaks and subsequent check valve slam, and blowdown isolation valve closure. These PUR/SGR forces are compared with the original design basis forces. Modeling techniques discussed include proper valve closure modeling, sonic velocity changes due to pipe material changes, and two phase flow effects. Lessons learned based on analyses done for several PWR PUR with SGR are presented. Lessons learned from these analyses include choosing the optimal replacement piping size and routing to improve system performance without resulting in excessive piping loads. (authors)

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

International Nuclear Information System (INIS)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

2016-01-01

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

Energy Technology Data Exchange (ETDEWEB)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

Design for whipping pipe impact on reinforced concrete panels

International Nuclear Information System (INIS)

Chen, C.C.; Gurbuz, O.

1984-01-01

This paper describes determination of local and overall effects on reinforced concrete panels due to whipping pipe impact in postulated pipe break events. Local damage includes the prediction of minimum concrete panel thickness required to prevent spalling from the back face of the target reinforced concrete panels. Evaluation of overall effect deals with the ductility ratio calculation for the target reinforced concrete panels. Design curves for determining the minimum panel thickness and the minimum reinforcement of reinforced concrete panels are presented in this paper for some cases commonly encountered in nuclear applications. The methodology and the results provided can be used to determine if an existing reinforced concrete wall is capable of resisting the whipping pipe impact, and consequently, if pipe whip restraints can be eliminated

Limit the effects of secondary circuit water or steam piping breaks in the reactor building

International Nuclear Information System (INIS)

Nachev, N.

2001-01-01

The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

Steel Fibers Reinforced Concrete Pipes - Experimental Tests and Numerical Simulation

Science.gov (United States)

Doru, Zdrenghea

2017-10-01

The paper presents in the first part a state of the art review of reinforced concrete pipes used in micro tunnelling realised through pipes jacking method and design methods for steel fibres reinforced concrete. In part two experimental tests are presented on inner pipes with diameters of 1410mm and 2200mm, and specimens (100x100x500mm) of reinforced concrete with metal fibres (35 kg / m3). In part two experimental tests are presented on pipes with inner diameters of 1410mm and 2200mm, and specimens (100x100x500mm) of reinforced concrete with steel fibres (35 kg / m3). The results obtained are analysed and are calculated residual flexural tensile strengths which characterise the post-cracking behaviour of steel fibres reinforced concrete. In the third part are presented numerical simulations of the tests of pipes and specimens. The model adopted for the pipes test was a three-dimensional model and loads considered were those obtained in experimental tests at reaching breaking forces. Tensile stresses determined were compared with mean flexural tensile strength. To validate tensile parameters of steel fibres reinforced concrete, experimental tests of the specimens were modelled with MIDAS program to reproduce the flexural breaking behaviour. To simulate post - cracking behaviour was used the method σ — ε based on the relationship stress - strain, according to RILEM TC 162-TDF. For the specimens tested were plotted F — δ diagrams, which have been superimposed for comparison with the similar diagrams of experimental tests. The comparison of experimental results with those obtained from numerical simulation leads to the following conclusions: - the maximum forces obtained by numerical calculation have higher values than the experimental values for the same tensile stresses; - forces corresponding of residual strengths have very similar values between the experimental and numerical calculations; - generally the numerical model estimates a breaking force greater

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

Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9: PRAISE computer code user's manual. Final report

International Nuclear Information System (INIS)

Lim, E.Y.

1981-08-01

The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. Crack propagation rates are governed by a Paris-type relationship with separate RMS cyclic stress intensity factors for the depth and length. Both uniform through the wall and radial gradient thermal stresses are included in the calculation of the stress intensity factors. The failure probabilities are estimated by applying Monte Carlo methods to simulate the life histories of the selected weld joint. In order to maximize computational efficiency, a stratified sampling procedure is used to select the initial crack size. Hydrostatic proof test, pre-service inspection, and in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor

Application of LBB to a nozzle-pipe interface

Energy Technology Data Exchange (ETDEWEB)

Yu, Y.J.; Sohn, G.H.; Kim, Y.J. [and others

1997-04-01

Typical LBB (Leak-Before-Break) analysis is performed for the highest stress location for each different type of material in the high energy pipe line. In most cases, the highest stress occurs at the nozzle and pipe interface location at the terminal end. The standard finite element analysis approach to calculate J-Integral values at the crack tip utilizes symmetry conditions when modeling near the nozzle as well as away from the nozzle region to minimize the model size and simplify the calculation of J-integral values at the crack tip. A factor of two is typically applied to the J-integral value to account for symmetric conditions. This simplified analysis can lead to conservative results especially for small diameter pipes where the asymmetry of the nozzle-pipe interface is ignored. The stiffness of the residual piping system and non-symmetries of geometry along with different material for the nozzle, safe end and pipe are usually omitted in current LBB methodology. In this paper, the effects of non-symmetries due to geometry and material at the pipe-nozzle interface are presented. Various LBB analyses are performed for a small diameter piping system to evaluate the effect a nozzle has on the J-integral calculation, crack opening area and crack stability. In addition, material differences between the nozzle and pipe are evaluated. Comparison is made between a pipe model and a nozzle-pipe interface model, and a LBB PED (Piping Evaluation Diagram) curve is developed to summarize the results for use by piping designers.

Risk analysis of in-service pressure piping containing defects

International Nuclear Information System (INIS)

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

2004-01-01

The reliability of pressure piping containing defects is important in engineering. The failure probability of pressure piping containing defects may be used as a guide to the most economic deployment of resources on maintenance, inspection and repair. This paper presents a probabilistic assessment methodology for in-service pressure piping containing defects, which is especially designed for programming. It is based on three assessment codes, BS 7910, R6 and SAPV-99, considering uncertainties in operating loadings, flaw sizes, material fracture toughness and flow stress. A general sampling computation method of stress intensity factor (SIF), in the form of the relationship between SIF and axial force and bending moment and torsion, is adopted. This relationship has been successfully used in developing software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), to assess planar and non-planar flaws. A numerical example is presented to illustrate the application of SAPP-2003 for calculating the failure probabilities of separate defects and for the assessed pressure piping

Applications of a fracture mechanics model of structural reliability to the effects of seismic events on reactor piping

International Nuclear Information System (INIS)

Harris, D.O.; Lim, E.Y.

1982-01-01

A fracture mechanics model of structural reliability is described. The model assumes that failure occurs due to the subcritical and catastrophic growth of as-fabricated defects. The material properties, stress history, number and dimensions of the initial cracks are treated as random variables. Crack growth is calculated using fracture mechanics principles. The model has been used to estimate the influence of earthquakes on the integrity of circumferential girth butt welds in the large (diameter greater than 30 in.) primary coolant system pipes of a commercial pressurized water reactor. In the absence of earthquakes, the probability of leaks and catastrophic double-ended guillotine breaks is estimated to be 10 -6 and 10 -12 per plant lifetime, respectively. These probabilities were only slightly increased by the occurrence of earthquakes. (author)

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.)

Leak before break analysis for cracking at multiple weld locations in BWR recirculation piping

International Nuclear Information System (INIS)

Zahoor, A.; Gamble, R.

1984-01-01

Periodically over the past decade, intergranular stress corrosion cracking (IGSCC) has been found in austenitic stainless steel piping at Boiling Water Reactor facilities. The effect of IGSCC on piping integrity has been evaluated previously in various BWR Owners Group and NRC studies. In these studies, the analyses were performed assuming the presence of a crack at a single weld location in the pipe run. The purpose of this investigation was to compare the leak rate and potential for unstable crack extension associated with a throughwall crack for the following two conditions in a BWR recirculation system: (1) the recirculation piping contains part through cracks at multiple weld locations and a single throughwall crack, and (2) the piping contains only a throughwall crack at one weld location. Two type BWRs were evaluated; namely, the ring header and five individual loop designs. The results from the analyses indicate that the potential for unstable crack extension at large bending loads, and leak rate at normal operation are not affected by the presence of part through cracks at multiple weld locations. The differences in the respective calculated L/sub eff/ and leak rates for the single and multiply cracked conditions are less than 2%

Evaluation of J-integral estimation scheme for flawed throughwall pipes

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1987-02-01

The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses.

Evaluation of J-integral estimation scheme for flawed throughwall pipes

International Nuclear Information System (INIS)

Zahoor, A.

1987-01-01

The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses. (orig.)

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.

Evaluation of flawed-pipe experiments: Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.

1986-11-01

The purpose of this work was to perform elastic plastic fracture mechanics evaluations of experimental data that have become available from the NRC Degraded Pipe Program, Phase II (DPII) and other NRC and EPRI sponsored programs. These evaluations were used to assess flaw evaluation procedures for austenitic and ferritic steel piping. The results also have application to leak before break fracture mechanics analysis. An improved relationship was developed for computing the J-Integral for pipes containing throughwall flaws and loaded in pure bending. The results from several DPII experiments were compared to predictions based on new J estimation scheme solutions for circumferential, finite length part-throughwall flaws in pipes with bending loading. Comparisons of experimental maximum loads with those predicted using procedures in Paragraph IWB-3640, Section XI of the ASME Code indicate that the Code flaw evaluation procedures and allowables for austenitic steel pipe are appropriate and conservative. However, the comparisons also indicate that the base metal Code allowable loads may be about 15 to 20% high for small diameter piping (less than 8-inch diameter) at allowable a/t larger than about 0.5. The work further indicates that there is justification for reducing the conservatism in IWB-3640 allowable flaw sizes and loads for austenitic steel pipe with submerged or shielded metal arc welds.

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

Fatigue analysis of HANARO primary cooling system piping

International Nuclear Information System (INIS)

Ryu, Jeong Soo

1998-05-01

A main form of piping failure which occurring leak before break (LBB) is fatigue failure. The fatigue analysis of HANARO primary cooling system (PCS) piping was performed. The PCS piping had been designed in accordance with ASME Class 3 for service conditions. However fatigue analysis is not required in Class 3. In this study the quantitative fatigue analysis was carried out according to ASME Class 1. The highest stress points which have the largest possibility of ASME class 1. The highest stress points which have the largest possibility of the fatigue were determined from the piping stress analysis for each subsection piping. The fatigue analysis was performed for 3 highest stress points, i.e., branch connection, anchor point and butt welding joint. After calculating the peak stress intensity range the fatigue usage factors were evaluated considering operating cycles and S-N curve. The cumulative usage factors for 3 highest stress points were much less than 1. The results show that the possibility of fatigue failure for PCS piping subjected to thermal expansion and seismic loads is very small. The structural integrity of the HANARO PCS piping for fatigue failure was proved to apply the LBB. (author). 11 tabs., 6 figs

Qualification of PHT piping of Indian 500 MW PHWR for LBB, using R-6 method

International Nuclear Information System (INIS)

Rastogi, Rohit; Bhasin, V.; Kushwaha, H.S.

1997-01-01

This document discusses the qualification of straight pipe portion of the primary heat transport (PHT) piping of Indian 500 MWe pressurised heavy water reactor (PHWR) for leak before break (LBB). The evaluation is done using R-6 [1] method. The results presented here are: the safety margins which exist on straight pipe components of main PHT piping of 500 MWe, under leakage size crack (LSC) and design basis accident loads; the sensitivity of safety margins with respect to different analysis parameters and the qualification of PHT piping for LBB based on criterion given by NUREG-1061 [2] and TECDOC-774 [3]. (author)

Experimental evaluation of J in cracked straight and curved pipes under bending

International Nuclear Information System (INIS)

Moulin, D.; Touboul, F.; Foucher, N.; Lebey, J.; Acker, D.

1989-01-01

An experimental program is being carried out at the CEA Saclay in collaboration with FRAMATOME and IPSN with a view to validate analysis methods applicable for evaluation of leak before break behavior in P.W.R. piping. A large experimental work was already performed in USA, Germany and Japan and cracked pipes made of stainless steel material under bending. The methods of analysis got same validations for straight pipes. However applicability to elbows and comparison with toughness values obtained on small specimens like CT specimens was not completely dealt with

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

Generic safety evaluation report regarding integrity of BWR scram system piping

International Nuclear Information System (INIS)

1981-08-01

Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service

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

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

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

Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

International Nuclear Information System (INIS)

Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

1982-01-01

One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

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

Causes of pipe ruptures in distribution lines. Evaluation of long-term observations in a metropolitan pipe network

Energy Technology Data Exchange (ETDEWEB)

Kottmann, A

1978-01-01

Pipe ruptures and their causes are examined from the viewpoints of pipe material, corrosion, traffic, internal pressure, air temperature, ground temperature, ground frost, gas or water temperature, and ground moisture level. The examination relies on 17 years of statistics (1958-74) from (1) Technische Werke der Stadt Stuttgart AG on 11,986 pipe ruptures and (2) German weather-service data on ground-moisture readings at depths down to 80 in. in the Stuttgart area. Faced with replacing up to 280 miles (450 km) of cast-iron gas-distribution lines that seemed extraordinarily prone to rupture (company records showed at least 20 breaks/month) after the conversion to natural gas, TWS authorized this study to determine the boundary conditions that make cast-iron pipe susceptible to fracture, thus minimizing the extent of the replacement program. The investigation showed that corrosion had only a slight effect upon cracking. No significant effect was found for any of the following: temperature-caused changes in material properties, internal pressure or pressure changes, fluctuations in gas temperature, changes in air temperature, and summertime changes in ground temperature. Stress loading by heavy traffic, however, doubled the fracture incidence.

Response of the steam generator VVER 1000 to a steam line break

International Nuclear Information System (INIS)

Novotny, J.; Novotny, J. Jr.

2003-01-01

Dynamic effects of a steam line break in the weld of the steam pipe and the steam collector on the steam generator system are analyzed. Modelling of a steam line break may concern two cases. The steam line without a restraint and the steam line protected by a whip restraint with viscous elements applied at the postulated break cross-section. The second case is considered. Programme SYSTUS offers a special element the stiffness and viscous damping coefficients of which may be defined as dependent on the relative displacement and velocity of its nodes respectively. A circumferential crack is simulated by a sudden decrease of longitudinal and lateral stiffness coefficients of these special SYSTUS elements to zero. The computation has shown that one can simulate the pipe to behave like completely broken during a time interval of 0,0001 s or less. These elements are used to model the whip restraint with viscous elements and viscous dampers of the GERB type as well. In the case of a whip restraint model the stiffness coefficient-displacement relation and damping coefficient - velocity relation are chosen to fit the given characteristics of the restraint. The special SYSTUS elements are used to constitute Maxwell elements modelling the elasto-plastic and viscous properties of the GERB dampers applied to the steam generator. It has been ascertained that a steam line break at the postulated weld crack between the steam pipe and the steam generator collector cannot endanger the integrity of the system even in a case of the absence of a whip restraint effect. (author)

Investigation on the reliability of expansion joint for piping with probabilistic method

International Nuclear Information System (INIS)

Ishii, Y.; Kambe, M.

1980-01-01

The reduction of the plant size is necessitated as one of the major targets in LMFBR design. Usually, piping work system is extensively used to absorb thermal expansion between two components anywhere. Besides above, expansion joint for piping seems to be attractive lately for the same object. This paper describes the significance of expansion joint with multiple boundaries, breakdown probability of expansion joint assembly and partly the bellows by introducing several hypothetical conditions in connection with piping. Also, an importance of in-service inspection (ISI) for expansion joint was discussed using a comparative table and probabilities on reliability from partly broken to full penetration. In conclusion, the expansion joint with ISI should be manufactured with excellent reliability in order to cope with piping work system; several conditions of the practical application for piping systems are suggested. (author)

Investigation on the reliability of expansion joint for piping with probabilistic method

Energy Technology Data Exchange (ETDEWEB)

Ishii, Y; Kambe, M

1980-02-01

The reduction of the plant size is necessitated as one of the major targets in LMFBR design. Usually, piping work system is extensively used to absorb thermal expansion between two components anywhere. Besides above, expansion joint for piping seems to be attractive lately for the same object. This paper describes the significance of expansion joint with multiple boundaries, breakdown probability of expansion joint assembly and partly the bellows by introducing several hypothetical conditions in connection with piping. Also, an importance of in-service inspection (ISI) for expansion joint was discussed using a comparative table and probabilities on reliability from partly broken to full penetration. In conclusion, the expansion joint with ISI should be manufactured with excellent reliability in order to cope with piping work system; several conditions of the practical application for piping systems are suggested. (author)

Investigation on the reliability of expansion joint for piping with probabilistic method

International Nuclear Information System (INIS)

Ishii, Yoichiro; Kambe, Mitsuru.

1979-11-01

The reduction of the plant size if necessitated as one of the major target in LMFBR design. Usually, piping work system is extensively used to absorb thermal expansion between two components anywhere. Besides above, expansion joint for piping seems to be attractive lately for the same object. This paper describes about the significance of expansion joint with multiple boundaries, breakdown probability of expansion joint assembly and partly the bellows by introducing several hypothetical conditions in connection with piping. Also, an importance of inservice inspection (ISI) for expansion joint was discussed using by comparative table and probabilities on reliability from partly broken to full penetration. In the conclusion, the expansion joint with ISI should be manufactured with excellent reliability in order to cope with piping work system, and several conditions of the practical application for piping systems are suggested. (author)

WHIPJET progress on piping restraint elimination at Beaver Valley - 2

International Nuclear Information System (INIS)

Server, W.L.; Szy Slow Ski, J.J.; Goldstein, N.A.

1986-01-01

Fracture mechanics technology has advanced to the point that an engineering approach using the concept of leak-before-break in lieu of postulating double-ended pipe rupture is now possible. An approach based upon this fracture mechanics technology, termed WHIPJET, is currently being applied to Beaver Valley Power Station, Unit 2 for Duquesne Light Company. The WHIPJET philosophy is simple, conservative, and provides defense-in-depth arguments for high energy piping throughout the balance-of-plant. Progress being made in applying WHIPJET to several lines is presented

Clogging of granular material in vertical pipes discharged at constant velocity

Directory of Open Access Journals (Sweden)

López-Rodríguez Diego

2017-01-01

Full Text Available We report an experimental study on the flow of spherical particles through a vertical pipe discharged at constant velocity by means of a conveyor belt placed at the bottom. For a pipe diameter 3.67 times the diameter of the particles, we observe the development of hanging arches that stop the flow as they are able to support the weight of the particles above them. We find that the distribution of times that it takes until a stable clog develops, decays exponentially. This is compatible with a clogging probability that remains constant during the discharge. We also observe that the probability of clogging along the pipe decreases with the height, i.e. most of the clogs are developed near the bottom. This spatial dependence may be attributed to different pressure values within the pipe which might also be related to a spontaneous development of an helical structure of the grains inside the pipe.

Recent evaluations of crack-opening-area in circumferentially cracked pipes

Energy Technology Data Exchange (ETDEWEB)

Rahman, S.; Brust, F.; Ghadiali, N.; Wilkowski, G.; Miura, N.

1997-04-01

Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology frequently requires calculation of leak rates. The leak rates depend on the crack-opening area of the through-wall crack in the pipe. In addition to LBB analyses which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks as are being evaluated in ASME Section XI. The objectives of this study were to review, evaluate, and refine current predictive models for performing crack-opening-area analyses of circumferentially cracked pipes. The results from twenty-five full-scale pipe fracture experiments, conducted in the Degraded Piping Program, the International Piping Integrity Research Group Program, and the Short Cracks in Piping and Piping Welds Program, were used to verify the analytical models. Standard statistical analyses were performed to assess used to verify the analytical models. Standard statistical analyses were performed to assess quantitatively the accuracy of the predictive models. The evaluation also involved finite element analyses for determining the crack-opening profile often needed to perform leak-rate calculations.

FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break

International Nuclear Information System (INIS)

NILSSON, Lars; GUSTAFSSON, Per-Ake; GUSTAFSON, Lennart; JANCZAK, Rajmund; OESTERLUNDH, Ingrid

1992-01-01

1 - Description of test facility: The FIX-II facility is a volume scaled 1:777 representation of a Swedish BWR with external pumps. The pressure vessel contains a 36 rod full length bundle and a spray condenser at the top to allow steady state operation. The downcomer, bypass channels and guide tube volumes are represented by external piping. The intact loop represents three of the four external reactor loops. The broken loop is constructed such that both guillotine breaks and split breaks may be simulated. The facility is equipped with ADS-simulation, but no ECCS injection are included. The FIX-II loop is also suited to investigate response of pump trips and MSIV closures in internal pump reactors. 2 - Description of test: Test 3025 simulates an intermediate size split break in one of the four main recirculation lines. The break area was 31 per cent of the scaled down pipe area of the reactor. The initial power of the 36-rod bundle was 3.38 MW, corresponding to the hot channel power of the reactor

Application of the cracked pipe element to creep crack growth prediction

Energy Technology Data Exchange (ETDEWEB)

Brochard, J.; Charras, T.

1997-04-01

The modification of a computer code for leak before break analysis is very briefly described. The CASTEM2000 code was developed for ductile fracture assessment of piping systems with postulated circumferential through-wall cracks under static or dynamic loading. The modification extends the capabilities of the cracked pipe element to the determination of fracture parameters under creep conditions (C*, {phi}c and {Delta}c). The model has the advantage of evaluating significant secondary effects, such as those from thermal loading.

Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

International Nuclear Information System (INIS)

Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

1987-01-01

From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 MPa were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. (orig./GL)

Development of residual stress prediction model in pipe weldment

Energy Technology Data Exchange (ETDEWEB)

Eom, Yun Yong; Lim, Se Young; Choi, Kang Hyeuk; Cho, Young Sam; Lim, Jae Hyuk [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2002-03-15

When Leak Before Break(LBB) concepts is applied to high energy piping of nuclear power plants, residual weld stresses is a important variable. The main purpose of his research is to develop the numerical model which can predict residual weld stresses. Firstly, basic theories were described which need to numerical analysis of welding parts. Before the analysis of pipe, welding of a flat plate was analyzed and compared. Appling the data of used pipes, thermal/mechanical analysis were accomplished and computed temperature gradient and residual stress distribution. For thermal analysis, proper heat flux was regarded as the heat source and convection/radiation heat transfer were considered at surfaces. The residual stresses were counted from the computed temperature gradient and they were compared and verified with a result of another research.

A Corrosion Risk Assessment Model for Underground Piping

Science.gov (United States)

Datta, Koushik; Fraser, Douglas R.

2009-01-01

The Pressure Systems Manager at NASA Ames Research Center (ARC) has embarked on a project to collect data and develop risk assessment models to support risk-informed decision making regarding future inspections of underground pipes at ARC. This paper shows progress in one area of this project - a corrosion risk assessment model for the underground high-pressure air distribution piping system at ARC. It consists of a Corrosion Model of pipe-segments, a Pipe Wrap Protection Model; and a Pipe Stress Model for a pipe segment. A Monte Carlo simulation of the combined models provides a distribution of the failure probabilities. Sensitivity study results show that the model uncertainty, or lack of knowledge, is the dominant contributor to the calculated unreliability of the underground piping system. As a result, the Pressure Systems Manager may consider investing resources specifically focused on reducing these uncertainties. Future work includes completing the data collection effort for the existing ground based pressure systems and applying the risk models to risk-based inspection strategies of the underground pipes at ARC.

Suitability of pipeline material for buried gas and water piping

Energy Technology Data Exchange (ETDEWEB)

Funk, R

1976-01-01

Following a brief review of the development of the individual pipe materials and their use in the field of gas and water supply, the various stressing possibilities are dealt with. The corrosion influences from inside and outside, the material specifically for internal and external insulation, as well as the stressing due to sediments, are particularly brought out in this connection. A few remarks on the pressure pipes made of ductile cast iron, steel, reinforced concrete, asbestos cement and plastics are followed by comparisons with representations on material parameters to be proved, safety factors, tensile and pressure resistance, breaking tension and stress-strain diagram, wall thicknesses, friction losses, reactions depending on the E. modulus and distribution of the single pipe materials in the gas and water supply.

Stresses in a curved pipe subject to an in-plane bending moment

International Nuclear Information System (INIS)

Hofmann, E.; Heeschen, U.

1979-01-01

The design of the KWU-primary component supports is mainly defined by the loads of the postulated pipe breaks. To estimate the maximum loading of a component support it is necessary to know the maximum in-plane bending moment (opening and closing) that can be transmitted by a pipe bend. Another reason for such information is that the displacements and distortions of the components cause higher stresses in elbows than in straight pipes. With a detailed knowledge of the deformation characteristic of a pipe bend an integrity analysis could be done without an expensive plastic system analysis. With this purpose in mind experiments were performed with straight pipes and pipe bends of different dimensions subject to in-plane bending moments. The experimental results give the ratio between the maximum transmittable moment of a pipe bend to that of a straight pipe or, the distortion of the end cross-sections and the flattening of the elbow cross-section. An attempt is made to derive simple expressions for estimating the behaviour at pipe elbows. Parallel to the experiments calculations were done for the straight pipe and elbow with a finite difference code with plastic capabilities. The results of the experiment and calculation are compared with the formulas of the ASME-Code section III subjection NB. (orig.)

Crack-opening area calculations for circumferential through-wall pipe cracks

Energy Technology Data Exchange (ETDEWEB)

Kishida, K.; Zahoor, A.

1988-08-01

This report describes the estimation schemes for crack opening displacement (COD) of a circumferential through-wall crack, then compares the COD predictions with pipe experimental data. Accurate predictions for COD are required to reliably predict the leak rate through a crack in leak-before-break applications.

Crack-opening area calculations for circumferential through-wall pipe cracks

International Nuclear Information System (INIS)

Kishida, K.; Zahoor, A.

1988-08-01

This report describes the estimation schemes for crack opening displacement (COD) of a circumferential through-wall crack, then compares the COD predictions with pipe experimental data. Accurate predictions for COD are required to reliably predict the leak rate through a crack in leak-before-break applications

Rupture hardware minimization in pressurized water reactor piping

International Nuclear Information System (INIS)

Mukherjee, S.K.; Ski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.F.; Quinones, D.F.; Server, W.L.

1989-01-01

For much of the high-energy piping in light reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied a Beaver Valley Power Station- Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferrutic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elemination

Analysis of NPP pipes and equipment damage in life time prolongation

International Nuclear Information System (INIS)

Tkachev, V.V.; Zheltukhin, K.K.

2008-01-01

Paper describes a procedure to calculate the probability of pipes and equipment failure taking account of both the service records of the structures under various conditions and their aging. The parameters characterizing applied loads, failures, as well as metal strength, mechanical and thermal properties serve as the arbitrary values used in the described procedure. Paper presents an example of the probability calculation of failure of the RBMK emergency feed pump recirculation pipes when their service life is prolonged [ru

Crack-tip constraint analyses and constraint-dependent LBB curves for circumferential through-wall cracked pipes

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.L.; Wang, G.Z., E-mail: gzwang@ecust.edu.cn; Xuan, F.Z.; Tu, S.T.

2015-04-15

Highlights: • Solution of constraint parameter τ* for through-wall cracked pipes has been obtained. • Constraint increases with increasing crack length and radius–thickness ratio of pipes. • Constraint-dependent LBB curve for through-wall cracked pipes has been constructed. • For increasing accuracy of LBB assessments, constraint effect should be considered. - Abstract: The leak-before-break (LBB) concept has been widely applied in the structural integrity assessments of pressured pipes in nuclear power plants. However, the crack-tip constraint effects in LBB analyses and designs cannot be incorporated. In this paper, by using three-dimensional finite element calculations, the modified load-independent T-stress constraint parameter τ* for circumferential through-wall cracked pipes with different geometries and crack sizes has been analyzed under different loading conditions, and the solutions of the crack-tip constraint parameter τ* have been obtained. Based on the τ* solutions and constraint-dependent J–R curves of a steel, the constraint-dependent LBB (leak-before-break) curves have been constructed. The results show that the constraint τ* increases with increasing crack length θ, mean radius R{sub m} and radius–thickness ratio R{sub m}/t of the pipes. In LBB analyses, the critical crack length calculated by the J–R curve of the standard high constraint specimen for pipes with shorter cracks is over-conservative, and the degree of conservatism increases with decreasing crack length θ, R{sub m} and R{sub m}/t. Therefore, the constraint-dependent LBB curves should be constructed to modify the over-conservatism and increase accuracy of LBB assessments.

On the behavior of pressurized pipings under excessive-stresses caused by earthquake loadings

International Nuclear Information System (INIS)

Udoguchi, Y.; Akino, K.; Shibata, H.

1975-01-01

Five types of breaking experiments on pipe elements and piping structures had been carried out from 1971 to 1973 by the technical sub-committee of the Japan Electric Association under the leadership taken by Y. Udoguchi, one of the authors. One of the fruitful results was to realize the guillotine-type rupture of pipe element on a shaking table. However, it was also shown that the margin for the design is enough, and allowable stresses under earthquake loading are obtained by modifying those of the Emergency Condition of the ASME Code. The experiments effected were as follows: straight pipe elements, curved pipes and T-branch pipe connections, made of both ferritic and austenitic steels, were subjected to repeated bending moment, torsional moment and combined under pressurized condition. The pressure corresponded to their design value, but the stresses caused by such moments exceeded over their allowable stress of the Faulted Condition of the ASME Code. The wave patterns were both sinusoidal and natural earthquake records

Reliability of CRBR primary piping: critique of stress-strength overlap method for cold-leg inlet downcomer

International Nuclear Information System (INIS)

Bari, R.A.; Buslik, A.J.; Papazoglou, I.A.

1976-04-01

A critique is presented of the strength-stress overlap method for the reliability of the CRBR primary heat transport system piping. The report addresses, in particular, the reliability assessment of WARD-D-0127 (Piping Integrity Status Report), which is part of the CRBR PSAR docket. It was found that the reliability assessment is extremely sensitive to the assumed shape for the probability density function for the strength (regarded as a random variable) of the cold-leg inlet downcomer section of the primary piping. Based on the rigorous Chebyschev inequality, it is shown that the piping failure probability is less than 10 -2 . On the other hand, it is shown that the failure probability can be much larger than approximately 10 -13 , the typical value put forth in WARD-D-0127

Probabilistic assessment of leak-before-break

International Nuclear Information System (INIS)

Bush, S.H.

1984-01-01

A summary of results illustrating what might be derived from a probabilistic risk assessment (PRA) study follows. The failure probabilities for larger sizes of nuclear piping are considered to be in the range of 10 -4 to 10 -6 per reactor-year (exclusive of intergranular stress corrosion cracking (IGSCC). Smaller pipe sizes, of lesser safety significance, have much higher failure rates. In BWRs, IGSCC can cause failure rates much higher than 10 -4 in piping 4 to 10 in. in size. Suggested failure mechanisms apply in most instances, exclusive of IGSCC. Catastrophic failures would appear more likely from operator error or design and construction errors (water hammer, improper handling of dynamic loads, and undetected fabrication defects) rather than conventional flaw initiation and growth by fatigue

Comparison of linear-elastic-plastic, and fully plastic failure models in the assessment of piping integrity

International Nuclear Information System (INIS)

Streit, R.D.

1981-01-01

The failure evaluation of Pressurized Water Reactor (PWR) primary coolant loop pipe is often based on a plastic limit load criterion; i.e., failure occurs when the stress on the pipe section exceeds the material flow stress. However, in addition the piping system must be safe against crack propagation at stresses less than those leading to plastic instability. In this paper, elastic, elastic-plastic, and fully-plastic failure models are evaluated, and the requirements for piping integrity based on these models are compared. The model yielding the 'more' critical criteria for the given geometry and loading conditions defines the appropriate failure criterion. The pipe geometry and loading used in this study was choosen based on an evaluation of a guillotine break in a PWR primary coolant loop. It is assumed that the piping may contain cracks. Since a deep circumferential crack, can lead to a guillotine pipe break without prior leaking and thus without warning it is the focus of the failure model comparison study. The hot leg pipe, a 29 in. I.D. by 2.5 in. wall thickness stainless pipe, was modeled in this investigation. Cracks up to 90% through the wall were considered. The loads considered in this evaluation result from the internal pressure, dead weight, and seismic stresses. For the case considered, the internal pressure contributes the most to the failure loading. The maximum moment stress due to the dead weight and seismic moments are simply added to the pressure stress. Thus, with the circumferential crack geometry and uniform pressure stress, the problem is axisymmetric. It is analyzed using NIKE2D--an implicit, finite deformation, finite element code for analyzing two-dimensional elastic-plastic problems. (orig./GL)

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

Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

International Nuclear Information System (INIS)

Isozaki, Toshikuni; Yano, Toshikazu; Miyazaki, Noriyuki; Kato, Rokuro; Kurihara, Ryoichi; Ueda, Shuzo; Miyazono, Shohachiro

1982-09-01

Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

xLPR - a probabilistic approach to piping integrity analysis

International Nuclear Information System (INIS)

Harrington, C.; Rudland, D.; Fyfitch, S.

2015-01-01

The xLPR Code is a probabilistic fracture mechanics (PFM) computational tool that can be used to quantitatively determine a best-estimate probability of failure with well characterized uncertainties for reactor coolant system components, beginning with the piping systems and including the effects of relevant active degradation mechanisms. The initial application planned for xLPR is somewhat narrowly focused on validating LBB (leak-before-break) compliance in PWSCC-susceptible systems such as coolant systems of PWRs. The xLPR code incorporates a set of deterministic models that represent the full range of physical phenomena necessary to evaluate both fatigue and PWSCC degradation modes from crack initiation through failure. These models are each implemented in a modular form and linked together by a probabilistic framework that contains the logic for xLPR execution, exercises the individual modules as required, and performs necessary administrative and bookkeeping functions. The completion of the first production version of the xLPR code in a fully documented, releasable condition is presently planned for spring 2015

Experimental and theoretical investigations on the behaviour of cracks in primary coolant piping

International Nuclear Information System (INIS)

Steinbuch, R.; Bartholome, G.; Felski, N.; Kastner, W.

1981-01-01

During the investigations of the government-sponsored R+D programs (RS 104 and RS 320) experimental and theoretical work has been performed to describe the leak before break behaviour and the extent of instable crack growth. The test pipes are 300 mm ID pipes made of 20MnMoNi55. Three of them had been welded to a pressure reservoir to simulate the situation of a real system of piping and components as related to hydrodynamics. The instrumentation of the specimen was designed to describe - temperature and pressure during failure - effect of reservoir on depressurisation - motion of the pipe - leakage area as function of time - crack arrest length. At two experiments the pressure dropped to saturation but in others for a short period the pressure was remarkably lower. (orig./GL)

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

Pipe rupture hardware minimization in pressurized water reactor system

International Nuclear Information System (INIS)

Mukherjee, S.K.; Szyslowski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.; Quinones, D.; Server, W.

1987-01-01

For much of the high energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station - Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in (152 mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination

Evaluation of piping fracture analysis method by benchmark study, 1

International Nuclear Information System (INIS)

Takahashi, Yukio; Kashima, Koichi; Kuwabara, Kazuo

1987-01-01

Importance of strength evaluation methods for cracked piping is growing with the progress of the rationalization of the nuclear piping system based on the leak-before-break concept. As an analytical tool, finite element method is principally used. To obtain the reliable solutions by the finite element programs, it is important to grasp the influences of various factors on the solutions. In this study, benchmark analysis is carried out for a stainless steel pipe with a circumferential through-wall crack subjected to four-point bending loading. Eight solutions obtained by using five finite element programs are compared with each other. Good agreement is obtained between the solutions on the deformation characteristics as well as fracture mechanics parameters. It is found through this study that the influence of the difference in the solution technique is generally small. (author)

Z → bb-bar probability and asymmetry in a model of dynamical electroweak symmetry breaking

International Nuclear Information System (INIS)

Arbuzov, B.A.; Osipov, M.Yu.

1997-01-01

The deviations from the standard model in the probability of Z → bb-bar decay and in the forward-backward asymmetry in the reaction e + e - → bb-bar are studied in the framework of the model of dynamical electroweak symmetry breaking, the basic point of which is the existence of a triple anomalous W-boson vertex in a region of momenta restricted by a cutoff. A set of equations for additional terms in the W b t-bar vertex is obtained and its solution to the process Z → bb-bar is applied. It is shown that it is possible to obtain a consistent description of both deviations, which is quite nontrivial because these effects are not simply correlated. The necessary value of the anomalous W interaction coupling, λ = -0.22 ± 0.01, is consistent with existing limitations and leads to definite predictions, e.g., for pair W production in e + e - collisions at LEP 200

Ductile fracture of circumferentially cracked type-304 stainless steel pipes in tension

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Norris, D.M.

1984-11-01

Circumferentially cracked pipes subjected to tensile load were analyzed for finite length and constant depth part-through cracks located at the inside of the pipe wall. The analysis postulated loads sufficient to cause net-section yielding of the flawed section. It was demonstrated that a propensity for predominantly radial growth exists for part-through cracks loaded in tension. This result is similar to the result for bend loading, except that bend loading causes more favorable conditions for wall breakthrough than tension loading. Numerical results were developed for 4-in. and 24-in-dia pipes. Safety margins for displacement controlled loads were described by a safety assessment diagram. This diagram defines a curve delineating leak from fracture in a space of nondimensional crack length and crack depth. 4-india schedule 80 Type-304 stainless steel pipes with length to radius ratio (L/R) of up to 100 exhibited leak-before-break behavior.

Ductile fracture of circumferentially cracked type-304 stainless steel pipes in tension

International Nuclear Information System (INIS)

Zahoor, A.; Norris, D.M.

1984-01-01

Circumferentially cracked pipes subjected to tensile load were analyzed for finite length and constant depth part-through cracks located at the inside of the pipe wall. The analysis postulated loads sufficient to cause net-section yielding of the flawed section. It was demonstrated that a propensity for predominantly radial growth exists for part-through cracks loaded in tension. This result is similar to the result for bend loading, except that bend loading causes more favorable conditions for wall breakthrough than tension loading. Numerical results were developed for 4-in. and 24-in-dia pipes. Safety margins for displacement controlled loads were described by a safety assessment diagram. This diagram defines a curve delineating leak from fracture in a space of nondimensional crack length and crack depth. 4-india schedule 80 Type-304 stainless steel pipes with length to radius ratio (L/R) of up to 100 exhibited leak-before-break behavior

Analysis of the failure performance of internally pressurized piping with surface flaws

International Nuclear Information System (INIS)

Iorio, A.F; Crespi, J.C.

1987-01-01

Due to frequent failures an Atucha I PHWR moderator circuit branch piping, made of stainless steel type AISI 347 (DIN 1.4550), studies have been made, involving the application of several fracture mechanics criteria, in order to determine the conditions of leak-before-break (L.BB) and the critical crack length of the piping. These studies lead to the conclusions that, for a straight pipe of outer diameter of 219 mm and 16 mm wall thickness, with a circumferential flaw and the principal stress being in the bending, the L.BB criteria are satisfied, being the critical crack length of the order of 400 mm. A better mechanical finishing and heat treatment was suggested in order to improve the resistance to crack initiation. (Author)

Ductile fracture of circumferentially cracked pipes subjected to bending loads

International Nuclear Information System (INIS)

Zahoor, A.; Kanninen, M.F.

1981-01-01

A plastic fracture mechanics methodology is presented for part-through cracks in pipes under bending. A previous analysis result on the behavior of part-through cracks in pipes is reviewed. Example quantitative results for the initiation and instability of radial growth of part-through cracks are presented and compared with the experimental data to demonstrate the applicability of the method. The analyses in our previous work are further developed to include the instability of circumferential growth of part-through cracks. Numerical results are then presented for a compliant piping system, under displacement controlled bending, which focus on (1) instability of radial growth (unstable wall breakthrough) and (2) instability of circumferential growth of the resulting throughthe-thickness crack. The combined results of the above two types of analyses are presented on a safety assessment diagram. This diagram defines a curve of critical combination of length and depth of part-through cracks which delineates leak from fracture. The effect of piping compliance on the leak-before-break assessment is discussed

Ductile fracture of circumferentially cracked pipes subjected to bending loads

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Kanninen, M.F.

1981-10-01

A plastic fracture mechanics methodology is presented for part-through cracks in pipes under bending. A previous analysis result on the behavior of part-through cracks in pipes is reviewed. Example quantitative results for the initiation and instability of radial growth of part-through cracks are presented and compared with the experimental data to demonstrate the applicability of the method. The analyses in our previous work are further developed to include the instability of circumferential growth of part-through cracks. Numerical results are then presented for a compliant piping system, under displacement controlled bending, which focus on (1) instability of radial growth (unstable wall breakthrough) and (2) instability of circumferential growth of the resulting throughthe-thickness crack. The combined results of the above two types of analyses are presented on a safety assessment diagram. This diagram defines a curve of critical combination of length and depth of part-through cracks which delineates leak from fracture. The effect of piping compliance on the leak-before-break assessment is discussed.

Further development of probabilistic analysis method for lifetime determination of piping and vessels. Final report; Weiterentwicklung probabilistischer Analysemethoden zur Lebensdauerbestimmung von Rohrleitungen und Behaeltern. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Heckmann, K.; Grebner, H.; Sievers, J.

2013-07-15

Within the framework of research project RS1196 the computer code PROST (Probabilistic Structure Calculation) for the quantitative evaluation of the structural reliability of pipe components has been further developed. Thereby models were provided and tested for the consideration of the damage mechanism 'stable crack growth' to determine leak and break probabilities in cylindrical structures of ferritic and austenitic reactor steels. These models are now additionally available to the model for the consideration of the damage mechanisms 'fatigue' and 'corrosion'. Moreover, a crack initiation model has been established supplementary to the treatment of initial cracks. Furthermore, the application range of the code was extended to the calculation of the growth of wall penetrating cracks. This is important for surface cracks growing until the formation of a stable leak. The calculation of the growth of the wall penetrating crack until break occurs improves the estimation of the break probability. For this purpose program modules were developed to be able to calculate stress intensity factors and critical crack lengths for wall penetrating cracks. In the frame of this work a restructuring of PROST was performed including possibilities to combine damage mechanisms during a calculation. Furthermore several additional fatigue crack growth laws were implemented. The implementation of methods to estimate leak areas and leak rates of wall penetrating cracks was completed by the inclusion of leak detection boundaries. The improved analysis methods were tested by calculation of cases treated already before. Furthermore comparative analyses have been performed for several tasks within the international activity BENCH-KJ. Altogether, the analyses show that with the provided flexible probabilistic analysis method quantitative determination of leak and break probabilities of a crack in a complex structure geometry under thermal-mechanical loading as

The Weakest Link : Spatial Variability in the Piping Failure Mechanism of Dikes

NARCIS (Netherlands)

Kanning, W.

2012-01-01

Piping is an important failure mechanism of flood defense structures. A dike fails due to piping when a head difference causes first the uplift of an inland blanket layer, and subsequently soil erosion due to a ground water flow. Spatial variability of subsoil parameters causes the probability of

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

International Nuclear Information System (INIS)

Verezemskij, V.G.

1986-01-01

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

Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

International Nuclear Information System (INIS)

Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

1987-01-01

From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in the reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 mPA were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. On account of the safety margins proved in the experiments, potential inaccuracies in theoretical structure analyses are recommended so as to be on the safe side. On the other hand, it appears that designing pipework with reference to elastic stress categories does not adequately take into account the actual reserves of the pipework material

Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs

Energy Technology Data Exchange (ETDEWEB)

Khaleel, Mohammad A.; Simonen, Fredric A.

2009-05-01

The work described in this report applies probabilistic structural mechanics models to predict the reliability of nuclear pressure boundary components. These same models are then applied to evaluate the effectiveness of alternative programs for inservice inspection to reduce these failure probabilities. Results of the calculations support the development and implementation of risk-informed inservice inspection of piping and vessels. Studies have specifically addressed the potential benefits of ultrasonic inspections to reduce failure probabilities associated with fatigue crack growth and stress-corrosion cracking. Parametric calculations were performed with the computer code pc-PRAISE to generate an extensive set of plots to cover a wide range of pipe wall thicknesses, cyclic operating stresses, and inspection strategies. The studies have also addressed critical inputs to fracture mechanics calculations such as the parameters that characterize the number and sizes of fabrication flaws in piping welds. Other calculations quantify uncertainties associated with the inputs calculations, the uncertainties in the fracture mechanics models, and the uncertainties in the resulting calculated failure probabilities. A final set of calculations address the effects of flaw sizing errors on the effectiveness of inservice inspection programs.

Application of break preclusion concept in German nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Roos, E. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Maier, V. [Bayernwerk AG, Muenchen (Germany); Nagel, G. [PraussenElektra AG, Hannover (Germany)] [and others

1997-04-01

The break preclusion concept is based on {open_quotes}KTA rules{close_quotes}, {open_quotes}RSK guidelines{close_quotes} and {open_quotes}Rahmenspeziflkation Basissicherheit{close_quotes}. These fundamental rules containing for example requirements on material, design, calculation, manufacturing and testing procedures are explained and the technical realisation is shown by means of examples. The proof of the quality of these piping systems can be executed by means of fracture mechanics calculations by showing that in every case the leakage monitoring system already detect cracks which are clearly smaller than the critical crack. Thus the leak before break behavior and the break preclusion concept is implicitly affirmed. In order to further diminish conservativities in the fracture mechanics procedures, specific research projects are executed which are explained in this contribution.

LHC Experimental Beam Pipe Upgrade during LS1

CERN Document Server

Lanza, G; Baglin, V; Chiggiato, P

2014-01-01

The LHC experimental beam pipes are being improved during the ongoing Long Shutdown 1 (LS1). Several vacuum chambers have been tested and validated before their installation inside the detectors. The validation tests include: leak tightness, ultimate vacuum pressure, material outgassing rate, and residual gas composition. NEG coatings are assessed by sticking probability measurement with the help of Monte Carlo simulations. In this paper the motivation for the beam pipe upgrade, the validation tests of the components and the results are presented and discussed.

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

Progress on the degraded piping program - Phase II. Battelle Columbus Division

International Nuclear Information System (INIS)

Wilkowski, Gery; Ahmad, J.; Barnes, C.; Brust, F.; Guerrieri, D.; Kramer, G.; Landow, M.; Marschall, C.; Nakagaki, M.; Papaspyropoulos; Scott, P.

1988-01-01

The overall objective of the Degraded Piping Program is to verify and improve simple estimation schemes to predict the fracture behavior of circumferentially cracked pipe. The program is limited to quasi-static fracture and cracks in straight pipe. There are a variety of materials, flaw geometries, pipe sizes, and loading conditions evaluated. The Degraded Piping Program,which has been extended for one more year, will supply results that provide a basis for regulatory decisions regard applications for leak-before-break (LBB) and In-service flaw assessment. The significance of our results are summarized relative to how they may affect regulatory technical needs. The scope of the work in The Degraded Piping Program includes both analytical and experimental efforts. The experimental efforts have concentrated on testing circumferentially cracked pipe at 550 F (288 C) under si-static loading. Many of the tasks within this program were undertaken with the objective of determining if any detailed efforts were needed. This is true for both the analytical and experimental efforts. i e of the tasks have been slightly expanded during the course of the gram, while others were found to be of lesser concern and further efforts in those areas were not pursued. The results of this summary include the efforts of the third year. These efforts have contributed considerably to the understanding of the application of elastic-plastic fracture mechanics to nuclear piping systems. Rather than listing the significant technical contributions, these contributions are summarized below in relation to their application to LBB analyses, in-service flaw assessment criteria, and (3) material characterization and unusual behavior of nuclear piping materials at light water reactor (LWR) temperatures

Fracture studies on stainless steel straight pipes under earthquake-type cyclic loading

International Nuclear Information System (INIS)

Raghava, G.; Vishnuvardhan, S.; Gandhi, P.; Vaze, K.K.

2014-01-01

In order to study the crack growth and cyclic fracture behaviour, which are required for realistic assessment of Leak Before Break (LBB) applicability, experimental investigations were carried out on straight pipes under quasi-crystal loading. Totally 13 pipes were tested; three were stainless steel welded (SSW) using conventional shielded metal arc welding (SMAW) technique and the remaining specimens were Narrow Gap Welded (NGW). The fracture tests were carried out under load control, displacement control and combination of the two; the pipes were subjected to different amplitudes of load or load-line displacement (LLD), which were decided based on the response of the pipes under monotonic loading. Cyclic tearing and crack growth studies on eight straight pipes of the same material reported earlier in published literature are also considered for studying the results and understanding the behaviour. Under load control, with almost equal load amplitude, the NGW pipe exhibited improved life in comparison with SMAW pipe when both are subjected to cyclic loading. The crack growth and tearing instability behaviour of the pipes were studied. The same were found to be different for load control, displacement control and combined control tests. Based in the load-controlled experimental results, material specific plot between cyclic load amplitude (as a percentage of maximum load carrying capacity of a specimen under monotonic fracture) and number of cycles to failure was obtained. The results indicate that the piping components subjected to quasi-cyclic loading may fail in very less number of cycles even when the load amplitude is sufficiently below the monotonic fracture/collapse load. These studies will be helpful in designing nuclear power plant (NPP) piping components subjected to earthquake-type cyclic loading. (author)

Failure behavior of a pipe system with a circumferentially orientated flaw - analytical and experimental investigations

International Nuclear Information System (INIS)

Mikkola, T.P.J.; Diem, H.; Blind, D.; Hunger, H.

1989-01-01

At the german HDR-test-facility a pipe failure experiment was performed at a fullsize feedwater piping system under operating conditions of T=240 0 C, p=10.6 MPa and with an elevated oxygen content in the pressure medium. The loading was internal pressure and a cyclic varying bending moment with an R-ratio of 0.5. The in form of a circumferentially orientated notch initially weakened piping system failed after a total number of 4773 loaded cycles with different frequencies in form of a small leak. The analyses of the fracture surface indicated the strongly growing influence of corrosion effects on the crack propagation rate with decreasing loading frequency. The cyclic crack growth and the leak-before-break behavior of the piping system could be explained on the basis of results of finite element calculations using ADINA-code. (orig.)

Probabilistic pipe fracture evaluations for applications to leak-rate detection

Energy Technology Data Exchange (ETDEWEB)

Rahman, S; Wilkowski, G; Ghadiali, N [Battelle Columbus Labs., OH (United States)

1993-12-31

Stochastic pipe fracture evaluations are conducted for applications to leak-rate detection. A state-of-the-art review was first conducted to evaluate the adequacy of current deterministic models for thermo-hydraulic and elastic-plastic fracture analyses. Then a new probabilistic model was developed with the above deterministic models for structural reliability analysis of cracked piping systems and statistical characterization of crack morphology parameters, material properties of pipe, and crack location. The proposed models are then applied for computing conditional probability of failure for various nuclear piping systems in BWR and PWR plants. The PRAISE code was not used, and the probabilistic model is based on modern methods of stochastic mechanics, computationally far superior to Monte Carlo and Stratified Sampling methods used in PRAISE. 10 refs., 9 figs., 1 tab.

Probabilistic pipe fracture evaluations for applications to leak-rate detection

International Nuclear Information System (INIS)

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

1992-01-01

Stochastic pipe fracture evaluations are conducted for applications to leak-rate detection. A state-of-the-art review was first conducted to evaluate the adequacy of current deterministic models for thermo-hydraulic and elastic-plastic fracture analyses. Then a new probabilistic model was developed with the above deterministic models for structural reliability analysis of cracked piping systems and statistical characterization of crack morphology parameters, material properties of pipe, and crack location. The proposed models are then applied for computing conditional probability of failure for various nuclear piping systems in BWR and PWR plants. The PRAISE code was not used, and the probabilistic model is based on modern methods of stochastic mechanics, computationally far superior to Monte Carlo and Stratified Sampling methods used in PRAISE. 10 refs., 9 figs., 1 tab

An elevator for locked drilling pipes

Energy Technology Data Exchange (ETDEWEB)

Gurbanov, R.S.; Abbasov, E.M.; Ismailov, A.A.; Mamedov, Yu.S.; Safarov, A.A.

1983-01-01

An elevator is proposed, which includes a body with a door. To reduce the probability of gas shows in a well with high speed lowering and lifting of the column of locked drilling pipes through providing the possibility of feeding a drilling mud in this case into the mine, the elevator is equipped with a pneumatic cylinder with a two way hollow rod, on one face of which a sealing element is mounted for sealing the drilling pipe and on the other, an adapter for feeding the drilling mud. The rod is linked with the sleeve of the pneumatic cylinder, which is rigidly linked with the body with the capability of axial movement without rotation.

Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

2000-01-01

The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

Squaring the Circle: Geometric Skewness and Symmetry Breaking for Passive Scalar Transport in Ducts and Pipes.

Science.gov (United States)

Aminian, Manuchehr; Bernardi, Francesca; Camassa, Roberto; McLaughlin, Richard M

2015-10-09

We study the role geometry plays in the emergence of asymmetries in diffusing passive scalars advected by pressure-driven flows in ducts and pipes of different aspect ratios. We uncover nonintuitive, multi-time-scale behavior gauged by a new statistic, which we term "geometric skewness" S^{G}, which measures instantaneously forming asymmetries at short times due to flow geometry. This signature distinguishes elliptical pipes of any aspect ratio, for which S^{G}=0, from rectangular ducts whose S^{G} is generically nonzero, and, interestingly, shows that a special duct of aspect ratio ≈0.53335 behaves like a circular pipe as its geometric skewness vanishes. Using a combination of exact solutions, novel short-time asymptotics, and Monte Carlo simulations, we establish the relevant time scales for plateaus and extrema in the evolution of the skewness and kurtosis for our class of geometries. For ducts limiting to channel geometries, we present new exact, single-series formulas for the first four moments on slices used to benchmark Monte Carlo simulations.

Development of crack shape: LBB methodology for cracked pipes

Energy Technology Data Exchange (ETDEWEB)

Moulin, D.; Chapuliot, S.; Drubay, B. [Commissariat a l Energie Atomique, Gif sur Yvette (France)

1997-04-01

For structures like vessels or pipes containing a fluid, the Leak-Before-Break (LBB) assessment requires to demonstrate that it is possible, during the lifetime of the component, to detect a rate of leakage due to a possible defect, the growth of which would result in a leak before-break of the component. This LBB assessment could be an important contribution to the overall structural integrity argument for many components. The aim of this paper is to review some practices used for LBB assessment and to describe how some new R & D results have been used to provide a simplified approach of fracture mechanics analysis and especially the evaluation of crack shape and size during the lifetime of the component.

Numerical evaluation of cracked pipes under dynamic loading

International Nuclear Information System (INIS)

Petit, M.; Jamet, P.

1989-01-01

In order to apply the leak-before-break concept to piping systems, the behavior of cracked pipes under dynamic, and especially seismic, loadings must be studied. A simple finite element model of a cracked pipe has been developed and implemented in the general purpose computer code CASTEM 2000. The model is a generalization of the approach proposed by Paris and Tada (1). Considered loads are bending moment and axial force (representing thermal expansion and internal pressure.) The elastic characteristics of the model are determined using the Zahoor formulae for the geometry-dependent factors. Owing to the material behabior plasticity must be taken into account. To represent the crack growth, the material is defined by two characteristic values: J 1c which is the level of energy corresponding to crack initiation and the tearing modulus, T, which governs the length of propagation of the crack. For dynamic loads, unilateral conditions are imposed to represent crack closure. The model has been used for the design of dynamic tests to be conducted on shaking tables. Test principle is briefly described and numerical results are presented. Finally evaluation of margin, due to plasticity, in comparison with the standard design procedure is made

Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

International Nuclear Information System (INIS)

Krajicek, J.E.

1977-01-01

This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

Development of two-phase flow along a large vertical pipe

International Nuclear Information System (INIS)

Dirk Lucas; Prasser, H.M.

2005-01-01

Full text of publication follows: To qualify CFD codes for two-phase flow simulations, closure laws describing the interaction between the phases are needed. Vertical pipe flow is a suitable object for studying the corresponding phenomena in case of dispersed bubbly flow. Here, the bubbles move under clear boundary conditions, resulting in a shear field of nearly constant structure where the bubbles rise for a comparatively long time. This allows to study the lateral motion of the bubbles in a shear flow as well as bubble coalescence and break-up by comparing gas volume fraction distributions and bubble size distributions at different heights. Very detailed data were obtained at the TOPFLOW facility of the Forschungszentrum Rossendorf using an advanced wire-mesh sensor. This sensor measures the instantaneous conductivity distribution over the pipe cross section. The high frequency of the measurement (2500 frames/s) allows the detection of single bubbles by a special evaluation procedure. Bubble size distributions, gas volume fraction distributions and also gas fraction distributions decomposed according to the bubble size are delivered as result of the evaluation procedure. The use of two sensors allows to measure the profile of the gas velocity. In previous works similar data for pipe of 51.2 mm inner diameter were used for the validation of non-drag bubble forces [1] and the evaluation of the influence of radial profiles on the development of the flow pattern [2]. First investigations on scaling effects were done using data obtained at a pipe with an inner diameter of 194 mm [3]. A constant distance between gas injection and measuring plane of L/D ∼ 40 was used. From a new test series now measurements are available for varying distances between the injection device and the wire-mesh sensor. This allows the evaluation of the development of the flow along the pipe. The data are used for the development and validation of mesoscale models for the forces acting on

Probabilistic based design rules for intersystem LOCAS in ABWR piping

International Nuclear Information System (INIS)

Ware, A.G.; Wesley, D.A.

1993-01-01

A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures

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.

Experimental study on dynamic pipe fracture in consideration of hydropower plant model

Directory of Open Access Journals (Sweden)

Kazumi Ishikawa

2009-12-01

Full Text Available In the case of sudden valve closure, water hammer creates the most powerful pressure and damage to pipeline systems. The best way to protect the pipeline system is to eliminate water hammer. The main reasons for water hammer occurrence are valve closure, high initial velocity, and static pressure. However, it is difficult to eliminate water hammer. Water hammer tends to occur when the valve is being closed. In this study, the pipe fracture caused by static water pressure, gradually increasing pressure, and suddenly increasing pressure were compared experimentally in a breaking PVC test pipe. The quasi-static zone, the dynamic zone, and the transition zone are defined through the results of those experiments, with consideration of the fracture patterns of test pipes and impulses. The maximum pressure results were used to design the pipeline even though it is in the dynamic zone.

Breaking of ocean surface waves

International Nuclear Information System (INIS)

Babanin, A.V.

2009-01-01

Wind-generated waves are the most prominent feature of the ocean surface, and so are breaking waves manifested by the appearance of sporadic whitecaps. Such breaking represents one of the most interesting and most challenging problems for both fluid mechanics and physical oceanography. It is an intermittent random process, very fast by comparison with other processes in the wave breaking on the water surface is not continuous, but its role in maintaining the energy balance within the continuous wind-wave field is critical. Ocean wave breaking also plays the primary role in the air-sea exchange of momentum, mass and heat, and it is of significant importance for ocean remote sensing, coastal and maritime engineering, navigation and other practical applications. Understanding the wave breaking its occurrence, the breaking rates and even ability to describe its onset has been hindered for decades by the strong non-linearity of the process, together with its irregular and ferocious nature. Recently, this knowledge has significantly advanced, and the review paper is an attempt to summarise the facts into a consistent, albeit still incomplete picture of the phenomenon. In the paper, variety of definitions related to the were breaking are discussed and formulated and methods for breaking detection and measurements are examined. Most of attention is dedicated to the research of wave breaking probability and severity. Experimental, observational, numerical and statistical approaches and their outcomes are reviewed. Present state of the wave-breaking research and knowledge is analysed and main outstanding problems are outlined (Authors)

Refinement and evaluation of crack-opening-area analyses for circumferential through-wall cracks in pipes

International Nuclear Information System (INIS)

Rahman, S.; Brust, F.; Ghadiali, N.; Krishnaswamy, P.; Wilkowski, G.; Choi, Y.H.; Moberg, F.; Brickstad, B.

1995-04-01

Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet impingement shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology frequently requires calculation of leak rates. These leak rates depend on the crack-opening area of a through-wall crack in the pipe. In addition to LBB analyses, which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks as are being evaluated in ASME Section 11. This study was requested by the NRC to review, evaluate, and refine current analytical models for crack-opening-area analyses of pipes with circumferential through-wall cracks. Twenty-five pipe experiments were analyzed to determine the accuracy of the predictive models. Several practical aspects of crack-opening such as; crack-face pressure, off-center cracks, restraint of pressure-induced bending, cracks in thickness transition regions, weld residual stresses, crack-morphology models, and thermal-hydraulic analysis, were also investigated. 140 refs., 105 figs., 41 tabs

Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

International Nuclear Information System (INIS)

2003-01-01

In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Influence of wetting effect at the outer surface of the pipe on increase in leak rate - experimental results and discussion

Energy Technology Data Exchange (ETDEWEB)

Isozaki, Toshikuni; Shibata, Katsuyuki

1997-04-01

Experimental and computed results applicable to Leak Before Break analysis are presented. The specific area of investigation is the effect of the temperature distribution changes due to wetting of the test pipe near the crack on the increase in the crack opening area and leak rate. Two 12-inch straight pipes subjected to both internal pressure and thermal load, but not to bending load, are modelled. The leak rate was found to be very susceptible to the metal temperature of the piping. In leak rate tests, therefore, it is recommended that temperature distribution be measured precisely for a wide area.

Pipe whip: a summary of the damage observed in BNL pipe-on-pipe impact tests

International Nuclear Information System (INIS)

Baum, M.R.

1987-01-01

This paper describes examples of the damage resulting from the impact of a whipping pipe on a nearby pressurised pipe. The work is a by-product of a study of the motion of a whipping pipe. The tests were conducted with small-diameter pipes mounted in rigid supports and hence the results are not directly applicable to large-scale plant applications where flexible support mountings are employed. The results illustrate the influence of whipping pipe energy, impact position and support type on the damage sustained by the target pipe. (author)

Technical report on the Piping Reliability Proving Tests at the Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1993-05-01

Japan Atomic Energy Research Institute (JAERI) conducts Piping Reliability Proving Tests from 1975 to 1992 based upon the contracts between JAERI and Science and Technology Agency of Japan (STA) under the auspices of the special account law for electric power development promotion. The purpose of these tests are to prove the structural reliability of the primary cooling piping constituting a part of the pressure boundary in the light water reactor power plants. The tests with large experimental facilities had ended already in 1990. Presently piping reliability analysis by the probabilistic fracture mechanics method is being done. Until now annual reports concerning the proving tests were produced and submitted to STA, whereas this report summarizes the test results done during these 16 years. Objectives of the piping reliability proving tests are to prove that the primary piping of the light water reactor (1) be reliable throughout the service period, (2) have no possibility of rupture, (3) bring no detrimental influence on the surrounding instrumentations or equipments near the break location even if it ruptured suddenly. To attain these objectives (i) pipe fatigue tests, (ii) unstable pipe fracture tests, (iii) pipe rupture tests and also the analyses by computer codes were done. After carrying out these tests, it is verified that the piping is reliable throughout the service period. The authors of this report are T. Isozaki, K. Shibata, S. Ueda, R. Kurihara, K. Onizawa and A. Kohsaka. The parts they wrote are shown in contents. (author)

Casing free district heating pipes; Mantelfria fjaerrvaermeroer

Energy Technology Data Exchange (ETDEWEB)

Saellberg, Sven-Erik; Nilsson, Stefan [Swedish National Testing and Research Inst., Goeteborg (Sweden)

2005-07-01

significant temperature movements have taken place. Consequently, no damage to the unprotected foam was seen. Neither could any signs be seen of deterioration of the PUR foam's mechanical and thermal insulation properties. Long-term effects from moisture exposure have though not been studied explicitly. With respect to production and transportation of the pipes, some sort of casing is probably necessary. But it seems possible to use casing free pipes, given that the foam is free from defects where liquid water may enter. As regards the extension of a moisture damaged zone, it is actually beneficial with no vapour tight casing present. The humidity may then diffuse outwards instead of along the pipe. A possible scenario would be to use some sort of simpler casing for pipes which are not to be laid in coarse grained backfill and are not subject to extensive temperature variations.

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

Ductile fracture behavior of 6-inch diameter type 304 stainless steel and STS 42 carbon steel piping containing a through-wall or part-through crack

International Nuclear Information System (INIS)

Shibata, Katsuyuki; Ohba, Toshihiro; Kawamura, Takaichi; Miyazono, Shohachiro; Kaneko, Tadashi; Yokoyama, Norio.

1986-05-01

The double ended guillotine break philosophy in the design base accident of the nuclear power plant is considered to be overly conservative from the view point of piping design. Through the past experiences and developments of the fabrication, inspection, and operation of nuclear power plants, it has been recognized that the Leak-Before-Break (LBB) concept can be justified in the LWR pressure boundary pipings. In order to verify the LBB concept, extensive experimental and theoretical works are being conducted in many countries. Furthermore, a revised piping design standard, in which LBB concept is introduced, is under preparation in Japan, U.S.A., and European countries. At JAERI, a research program to investigate the unstable ductile fracture behavior of LWR piping under bending load has been carried out as a part of the LBB verification researches since 1983. This report summarizes the result of the ductile fracture tests conducted at room temperature in 1983 and 84. The 6-inch diameter pipes of type 304 stainless steel and STS 42 carbon steel pipe with a through-wall or part-through crack were tested under bending load with low or high compliance condition at room temperature. Pipe fracture data were obtained from the test as regards to load- displacement curve, crack extension, net section stress, J-resistance curve, and so on. Besides, the influence of the compliance on the fracture behavior was examined. Discussions are performed on the ductile pipe fracture criterion, flaw evaluation criterion, and LBB evaluation method. (author)

Experimental study on size effect of siphon-breaking hole in the real-scaled reactor pool

International Nuclear Information System (INIS)

Kang, Soon Ho; Ahn, Ho Seon; Kim, Ji Min; Kim, Moo Hwan; Lee, Kwon Yeong; Seo, Kyoung Woo; Chi, Dae Young

2012-01-01

A rupture in the primary piping of a cooling system with a heat source or in a research reactor could lead to a loss-of-coolant accident (LOCA). However, if the water level of the reactor pool could be sustained and a reactor scram follows, the heat source could be cooled by natural convection, and significant accidents could be avoided. When a piping-system rupture accident occurs, the coolant starts to siphon out of the reactor pool until the pressure head between the inlet and outlet is removed or the siphon flow is interrupted. Therefore, a siphon-breaker mechanism can be adopted as a passive safety device to maintain the reactor water level. The gas entrainment is used to block the continuous loss of coolant by interrupting the siphon flow. Siphon breaking is complicated due to the transient, turbulent, two-phase flow mode, so suitable models or correlations that describe this phenomenon do not exist, and no general analysis been developed. Previous researchers have conducted experiments and numerical simulations to design a siphon breaker to meet their needs. Previous research on siphon breaking has not been conducted systemically, and no literature exists, even though the topic is greatly concerned with hydraulic safety. In this study, siphon-breaking holes were used as siphon breakers, and their performance was evaluated by the residual water quantity. Flow visualization was conducted to interpret the siphon-breaking phenomenon

Pipe clamp effects on thin-walled pipe design

International Nuclear Information System (INIS)

Lindquist, M.R.

1980-01-01

Clamp induced stresses in FFTF piping are sufficiently large to require structural assessment. The basic principles and procedures used in analyzing FFTF piping at clamp support locations for compliance with ASME Code rules are given. Typical results from a three-dimensional shell finite element pipe model with clamp loads applied over the clamp/pipe contact area are shown. Analyses performed to categorize clamp induced piping loads as primary or secondary in nature are described. The ELCLAMP Computer Code, which performs analyses at clamp locations combining clamp induced stresses with stresses from overall piping system loads, is discussed. Grouping and enveloping methods to reduce the number of individual clamp locations requiring analysis are described

Quality Determination of Pipe-Borne Water in Sokoto Metropolis ...

African Journals Online (AJOL)

The quality of the pipe-borne water supplied to Sokoto metropolis was determined in this study. The total bacterial count was carried out using surface plating method of inoculation. The coliforms were enumerated using multiple tube fermentation technique (Most Probable Number Method). Some physicochemical ...

Elasto-plastic finite element analysis of axial surface crack in PHT piping of 500 MWe PHWR

International Nuclear Information System (INIS)

Chawla, D.S.; Bhate, S.R.; Kushwaha, H.S.; Mahajan, S.C.

1994-01-01

The leak before break (LBB) approach in nuclear piping design envisages demonstrating that the pressurized pipe with a postulated flaw will leak at a detectable rate leading to corrective action well before catastrophic rupture would occur. This requires analysis of cracked pipe to study the crack growth and its stability. This report presents the behaviour of a surface crack in the wall of a thick primary heat transport (PHT) pipe of 500 MWe Indian PHWR. The line spring model (LSM) finite element is used to model the flawed pipe geometry. The variation of crack driving force (J-integral) across the crack front has been presented. The influence of crack geometry factors such as depth, shape, aspect ratio, and loading on peak values of J-integral as well as crack mouth opening displacement has been studied. Several crack shapes have been used to study the shape influence. The results are presented in dimensionless form so as to widen their applicability. The accuracy of the results is validated by comparison with results available in open literature. (author). 47 refs., 8 figs

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.

Advanced management of pipe wall thinning based on prediction-monitor fusion

International Nuclear Information System (INIS)

Kojima, Fumio; Uchida, Shunsuke

2012-01-01

This article is concerned with pipe wall thinning management system by means of hybrid use of simulation and monitoring. First, the computer-aided simulation for predicting wear rate of piping system is developed based on elucidation of thinning mechanism such as flow-accelerated corrosion (FAC). The accurate prediction of wear rate allows us the useful information on region of interest of inspection. Secondly, several monitoring methods are considered in accordance with interest of inspection. Thirdly, probability of detection (POD) is considered for the reliability of inspection data. The final part of this article is devoted to how to improve safety performance under the hybrid use of predicting and monitoring on the proposed pipe wall management. (author)

Evaluation of throughwall crack pipes under displacement controlled loading

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1987-02-01

Tearing modulus solutions are developed for flawed throughwall pipes subjected to displacement controlled loading. Two cases of loading were considered: (1) a displacement controlled bending loading, and (2) a displacement controlled axial tension loading. A revised version of the EPRI J-integral estimation scheme is used in the development of the solutions. These solutions can be used for the entire range of elastic-plastic loading, from linear elastic, contained yielding, to large scale yielding of the crack section. Experimental data from pipes in bending were used to assess the accuracy of the compliant loading solutions. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. These solutions are shown to have good accuracy when used to predict the experimental results. The methodology and procedure can also be applied to part-throughwall cracks. These solutions have application to the leak before break fracture mechanics analyses.

Evaluation of throughwall crack pipes under displacement controlled loading

International Nuclear Information System (INIS)

Zahoor, A.

1987-01-01

Tearing modulus solutions are developed for flawed throughwall pipes subjected to displacement controlled loading. Two cases of loading were considered: (1) a displacement controlled bending loading, and (2) a displacement controlled axial tension loading. A revised version of the EPRI J-integral estimation scheme is used in the development of the solutions. These solutions can be used for the entire range of elastic-plastic loading, from linear elastic, contained yielding, to large scale yielding of the crack section. Experimental data from pipes in bending were used to assess the accuracy of the compliant loading solutions. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. These solutions are shown to have good accuracy when used to predict the experimental results. The methodology and procedure can also be applied to part-throughwall cracks. These solutions have application to the leak before break fracture mechanics analyses. (orig.)

Evaluation of effective J-integral value for 3-D TWC pipe in ABAQUS code

International Nuclear Information System (INIS)

Yang, J. S.; You, K. W.; Sung, K. B.; Jung, W. T.; Kim, B. N.

1999-01-01

This paper suggests a simple method to estimate the effective J-integral values in applying Leak-Before-Break (LBB) technology to nuclear piping system. In this paper, the effective J-integral estimates were calculated using energy domain integral approach with ABAQUS computer program. In this case, there existed a apparent variation of J-integral values along the crack line through the thickness of pipe. For this reason, several case studies have been performed to evaluate the effective J-integral value. From the results, it was concluded that the simple method suggested in this paper can be effectively used in estimating the effective J-integral value

Simulation of boiling flow in evaporator of separate type heat pipe with low heat flux

International Nuclear Information System (INIS)

Kuang, Y.W.; Wang, Wen; Zhuan, Rui; Yi, C.C.

2015-01-01

Highlights: • A boiling flow model in a separate type heat pipe with 65 mm diameter tube. • Nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux. • The two-phase heat transfer coefficient is less sensitive to the total mass flux. - Abstract: The separate type heat pipe heat exchanger is considered to be a potential selection for developing passive cooling spent fuel pool – for the passive pressurized water reactor. This paper simulates the boiling flow behavior in the evaporator of separate type heat pipe, consisting of a bundle of tubes of inner diameter 65 mm. It displays two-phase characteristic in the evaporation section of the heat pipe working in low heat flux. In this study, the two-phase flow model in the evaporation section of the separate type heat pipe is presented. The volume of fluid (VOF) model is used to consider the interaction between the ammonia gas and liquid. The flow patterns and flow behaviors are studied and the agitated bubbly flow, churn bubbly flow are obtained, the slug bubble is likely to break into churn slug or churn froth flow. In addition, study on the heat transfer coefficients indicates that the nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux, with the heat transfer coefficient being less sensitive to the total mass flux

The Coulomb break-up of 9Be

International Nuclear Information System (INIS)

Macdonald, E.W.; Shotter, A.C.; Branford, D.; Rahighi, J.; Davinson, T.; Davis, N.J.

1992-01-01

Kinematically complete data is presented on the break-up reaction 120 Sn( 9 Be, 8 Be g.s +n) 120 Sn g.s. at E beam =90 MeV for several scattering angles inside the grazing angle. These data are compared with the predictions of a Coulomb break-up model. It is shown that the data can be understood in terms of the Coulomb model provided some account is taken of the interactions of the break-up fragments with the target. Analysis of the 9 Be break-up data, using radio-isotope measurements of the 9 Be(γ, n) cross-section, indicates that for this photo-disintegration reaction there is probably a significant direct component to the threshold cross-section, in addition to a threshold resonance at 1.69 MeV. (orig.)

Steam line break analysis in CAREM-25 reactor

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

2000-01-01

The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

Steam Line Break Analysis in CAREM-25 Reactor

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo; Vertullo, Alicia; Garcia, A; Schlamp, Miguel

2000-01-01

The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model.The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator.As a consequence and due to reactor features the core power is also increased.As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided.Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System.In all the sequences the DNBR and CPR remain above the minimum safety values established by design.Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

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

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)

ADIMEW: Fracture assessment and testing of an aged dissimilar metal weld pipe assembly

International Nuclear Information System (INIS)

Wintle, J.B.; Hayes, B.; Goldthorpe, M.R.

2004-01-01

ADIMEW (Assessment of Aged Piping Dissimilar Metal Weld Integrity) was a three-year collaborative research programme carried out under the EC 5th Framework Programme. The objective of the study was to advance the understanding of the behaviour and safety assessment of defects in dissimilar metal welds between pipes representative of those found in nuclear power plant. ADIMEW studied and compared different methods for predicting the behaviour of defects located near the fusion boundaries of dissimilar metal welds typically used to join sections of austenitic and ferritic piping operating at high temperature. Assessment of such defects is complicated by issues that include: severe mis-match of yield strength of the constituent parent and weld metals, strong gradients of material properties, the presence of welding residual stresses and mixed mode loading of the defect. The study includes the measurement of material properties and residual stresses, predictive engineering analysis and validation by means of a large-scale test. The particular component studied was a 453mm diameter pipe that joins a section of type A508 Class 3 ferritic pipe to a section of type 316L austenitic pipe by means of a type 308 austenitic weld with type 308/309L buttering laid on the ferritic pipe. A circumferential, surface-breaking defect was cut using electro discharge machining into the 308L/309L weld buttering layer parallel to the fusion line. The test pipe was subjected to four-point bending to promote ductile tearing of the defect. This paper presents the results of TWI contributions to ADIMEW including: fracture toughness testing, residual stress measurements and assessments of the ADIMEW test using elastic-plastic, cracked body, finite element analysis. (orig.)

Prediction of LOCA Break Size Using CFNN

Energy Technology Data Exchange (ETDEWEB)

Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Kim, Dong Yeong; Na, Man Gyun [Chosun University Gwangju (Korea, Republic of)

2016-05-15

The NPPs have the emergency core cooling system (ECCS) such as a safety injection system. The ECCS may not function properly in case of the small break size due to a slight change of pressure in the pipe. If the coolant is not supplied by ECCS, the reactor core will melt. Therefore, the meltdown of reactor core have to be prevented by appropriate accident management through the prediction of LOCA break size in advance. This study presents the prediction of LOCA break size using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model is a data-based method that requires data for its development and verification. The data were obtained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using MAAP code, because real severe accident data cannot be obtained from actual NPP accidents. The CFNN model has been designed to rapidly predict the LOCA break size in LOCA situations. The CFNN model was trained by using the training data set and checked by using test data set. These data sets were obtained using MAAP code for OPR1000 reactor. The performance results of the CFNN model show that the RMS error decreases as the stage number of the CFNN model increases. In addition, the performance result of the CFNN model presents that the RMS error level is below 4%.

Analytical model of impedance in elliptical beam pipes

CERN Document Server

Pesah, Arthur Chalom

2017-01-01

Beam instabilities are among the main limitations in building higher intensity accelerators. Having a good impedance model for every accelerators is necessary in order to build components that minimize the probability of instabilities caused by the interaction beam-environment and to understand what piece to change in case of intensity increasing. Most of accelerator components have their impedance simulated with finite elements method (using softwares like CST Studio), but simple components such as circular or flat pipes are modeled analytically, with a decreasing computation time and an increasing precision compared to their simulated model. Elliptical beam pipes, while being a simple component present in some accelerators, still misses a good analytical model working for the hole range of velocities and frequencies. In this report, we present a general framework to study the impedance of elliptical pipes analytically. We developed a model for both longitudinal and transverse impedance, first in the case of...

DNA double-strand-break complexity levels and their possible contributions to the probability for error-prone processing and repair pathway choice.

Science.gov (United States)

Schipler, Agnes; Iliakis, George

2013-09-01

Although the DNA double-strand break (DSB) is defined as a rupture in the double-stranded DNA molecule that can occur without chemical modification in any of the constituent building blocks, it is recognized that this form is restricted to enzyme-induced DSBs. DSBs generated by physical or chemical agents can include at the break site a spectrum of base alterations (lesions). The nature and number of such chemical alterations define the complexity of the DSB and are considered putative determinants for repair pathway choice and the probability that errors will occur during this processing. As the pathways engaged in DSB processing show distinct and frequently inherent propensities for errors, pathway choice also defines the error-levels cells opt to accept. Here, we present a classification of DSBs on the basis of increasing complexity and discuss how complexity may affect processing, as well as how it may cause lethal or carcinogenic processing errors. By critically analyzing the characteristics of DSB repair pathways, we suggest that all repair pathways can in principle remove lesions clustering at the DSB but are likely to fail when they encounter clusters of DSBs that cause a local form of chromothripsis. In the same framework, we also analyze the rational of DSB repair pathway choice.

Two-phase flow through small branches in a horizontal pipe with stratified flow

International Nuclear Information System (INIS)

Smoglie, C.

1985-02-01

In the field of reactor safety the occurrence of a small break in a horizontal primary coolant pipe is of great importance. This report presents the description and results of experiments designed to determine the mass flow rate and quality through a small break at the bottom, the top or the side of a main pipe with stratified gas-liquid flow. If the interface level is far below (above) the branch, only single-phase gas (liquid) flow enters the branch. For smaller distances the interface is locally deformed because of the pressure decrease due to the fluid acceleration near the branch inlet (Bernoulli effect) and liquid (gas) can be entrained. This report contains photographs illustrating the flow phenomena as well as a general correlation to determine the beginning of entrainment. Results are presented on the branch mass flow rate and quality as a function of a normalized distance between the interface and the branch inlet. A model was developed which enables to predict the branch quality and mass flux. Results from air-water flow through horizontal branches, were extrapolated for steam water flow at high pressure with critical branch mass flux. (orig./HS) [de

Users manual on database of the Piping Reliability Proving Tests at the Japan Atomic Energy Research Institute

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

Japan Atomic Energy Research Institute(JAERI) conducted Piping Reliability Proving Tests from 1975 to 1992 based upon the contracts between JAERI and Science and Technology Agency of Japan under the auspices of the special account law for electric power development promotion. The purposes of those tests are to prove the structural reliability of the primary cooling piping constituting a part of the pressure boundary in the water reactor power plants. The tests with large experimental facilities had ended already in 1990. After that piping reliability analysis by the probabilistic method followed until 1992. This report describes the users manual on databases about the test results using the large experimental facilities. Objectives of the piping reliability proving tests are to prove that the primary piping of the water reactor (1) be reliable throughout the service period, (2) have no possibility of rupture, (3) bring no detrimental influence on the surrounding instrumentations or equipments near the break location. The research activities using large scale piping test facilities are described. The present report does the database about the test results pairing the former report. With these two reports, all the feature of Piping Reliability Proving Tests is made clear. Briefings of the tests are described also written in Japanese or English. (author)

Proceedings of a specialist meeting on leak before break in reactor piping and vessels

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

This Specialist Meeting was organised by EDF, Framatome and CEA with the participation of SFEN, and it was sponsored jointly by the CEC DG XI, Nuclear Electric, IAEA, US NRC, and by the Principal Working Group 3 (PWG-3) on Reactor Component Integrity of the NEA CSNI. The activities of PWG-3 fall into three main areas: Non-Destructive Examination (NDE), fracture analysis and aging/materials degradation. In fracture analysis, the activities are organised by the Fracture Analysis Group, and include the round robins on Fracture Analysis of Large Scale International Reference Experiments (FALSIRE). The topic of the workshop falls mainly into the second area of fracture analysis. The objective of the meeting was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. The formal proceedings of the meeting were published by US NRC as a NUREG report (NUREG/CP--0155). This includes the final versions of papers

Proceedings of a specialist meeting on leak before break in reactor piping and vessels

International Nuclear Information System (INIS)

1996-01-01

This Specialist Meeting was organised by EDF, Framatome and CEA with the participation of SFEN, and it was sponsored jointly by the CEC DG XI, Nuclear Electric, IAEA, US NRC, and by the Principal Working Group 3 (PWG-3) on Reactor Component Integrity of the NEA CSNI. The activities of PWG-3 fall into three main areas: Non-Destructive Examination (NDE), fracture analysis and aging/materials degradation. In fracture analysis, the activities are organised by the Fracture Analysis Group, and include the round robins on Fracture Analysis of Large Scale International Reference Experiments (FALSIRE). The topic of the workshop falls mainly into the second area of fracture analysis. The objective of the meeting was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. The formal proceedings of the meeting were published by US NRC as a NUREG report (NUREG/CP--0155). This includes the final versions of papers

PPOOLEX experiments with a modified blowdown pipe outlet

International Nuclear Information System (INIS)

Laine, J.; Puustinen, M.; Raesaenen, A.

2009-08-01

This report summarizes the results of the experiments with a modified blowdown pipe outlet carried out with the PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through a vertical DN200 blowdown pipe to the condensation pool. Four reference experiments with a straight pipe and ten with the Forsmark type collar were carried out. The main purpose of the experiment series was to study the effect of a blowdown pipe outlet collar design on loads caused by chugging phenomena (rapid condensation) while steam is discharged into the condensation pool. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. During the experiments the initial temperature level of the condensation pool water was either 20-25 or 50-55 deg. C. The steam flow rate varied from 400 to 1200 g/s and the temperature of incoming steam from 142 to 185 deg. C. In the experiments with 20-25 deg. C pool water, even 10 times higher pressure pulses were measured inside the blowdown pipe in the case of the straight pipe than with the collar. In this respect, the collar design worked as planned and removed the high pressure spikes from the blowdown pipe. Meanwhile, there seemed to be no suppressing effect on the loads due to the collar in the pool side in this temperature range. Registered loads in the pool were approximately in the same range (or even a little higher) with the collar as with the straight pipe. In the experiments with 50-55 deg. C pool water no high pressure pulses were measured inside the blowdown pipe either with the straight pipe or with the collar. In this case, more of the suppressing effect is probably due to the warmer pool water than due to the modified pipe outlet. It has been observed already in the earlier experiments with a straight pipe in the POOLEX and PPOOLEX facilities that warm pool water has a diminishing effect on

PPOOLEX experiments with a modified blowdown pipe outlet

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

2009-08-15

This report summarizes the results of the experiments with a modified blowdown pipe outlet carried out with the PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through a vertical DN200 blowdown pipe to the condensation pool. Four reference experiments with a straight pipe and ten with the Forsmark type collar were carried out. The main purpose of the experiment series was to study the effect of a blowdown pipe outlet collar design on loads caused by chugging phenomena (rapid condensation) while steam is discharged into the condensation pool. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. During the experiments the initial temperature level of the condensation pool water was either 20-25 or 50-55 deg. C. The steam flow rate varied from 400 to 1200 g/s and the temperature of incoming steam from 142 to 185 deg. C. In the experiments with 20-25 deg. C pool water, even 10 times higher pressure pulses were measured inside the blowdown pipe in the case of the straight pipe than with the collar. In this respect, the collar design worked as planned and removed the high pressure spikes from the blowdown pipe. Meanwhile, there seemed to be no suppressing effect on the loads due to the collar in the pool side in this temperature range. Registered loads in the pool were approximately in the same range (or even a little higher) with the collar as with the straight pipe. In the experiments with 50-55 deg. C pool water no high pressure pulses were measured inside the blowdown pipe either with the straight pipe or with the collar. In this case, more of the suppressing effect is probably due to the warmer pool water than due to the modified pipe outlet. It has been observed already in the earlier experiments with a straight pipe in the POOLEX and PPOOLEX facilities that warm pool water has a diminishing effect on

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Prediction of Counter-Current Flow Limitation at Hot Leg Pipe During a Small-Break Loca

Energy Technology Data Exchange (ETDEWEB)

Jeong, H.Y. [Korea Electric Power Research Institute, Taejeon (Korea)

2001-07-01

The possibility of hot leg flooding during reflux condensation cooling after a small-break loss-of-coolant accident in a nuclear power plant is evaluated. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The effect of initial water level to counter-current flow limitation is taken into account. It is predicted that the hot leg flooding is precluded when all steam generators are available for heat removal. It is also shown the both hot leg flooding and SG flooding are possible under the operation of one steam generators. Therefore, it can be said that the occurrence of hot leg flooding under reflux condensation cooling is possible when the number of steam generators available for heat removal is limited. (author). 15 refs., 15 figs., 3 tabs.

The Second International Piping Integrity Research Group (IPIRG-2) program. Final report, October 1991--April 1996

International Nuclear Information System (INIS)

Hopper, A.; Wilowski, G.; Scott, P.; Olson, R.

1997-03-01

The IPIRG-2 program was an international group program managed by the US NRC and funded by organizations from 15 nations. The emphasis of the IPIRG-2 program was the development of data to verify fracture analyses for cracked pipes and fittings subjected to dynamic/cyclic load histories typical of seismic events. The scope included: (1) the study of more complex dynamic/cyclic load histories, i.e., multi-frequency, variable amplitude, simulated seismic excitations, than those considered in the IPIRG-1 program, (2) crack sizes more typical of those considered in Leak-Before-Break (LBB) and in-service flaw evaluations, (3) through-wall-cracked pipe experiments which can be used to validate LBB-type fracture analyses, (4) cracks in and around pipe fittings, such as elbows, and (5) laboratory specimen and separate effect pipe experiments to provide better insight into the effects of dynamic and cyclic load histories. Also undertaken were an uncertainty analysis to identify the issues most important for LBB or in-service flaw evaluations, updating computer codes and databases, the development and conduct of a series of round-robin analyses, and analyst's group meetings to provide a forum for nuclear piping experts from around the world to exchange information on the subject of pipe fracture technology. 17 refs., 104 figs., 41 tabs

Investigation of straitified and countercurrent flows in horizontal piping during a loss-of-coolant accident

International Nuclear Information System (INIS)

Bourteele, J.P.

1980-06-01

The ECTHOR program consists in a loop having as objective to study the flow regimes in horizontal pipings (stratification, countercurrent flows) in conditions representative of small break transients within commercial PWR. The ECTHOR tests are in process. Experimental results are already available and are presented in this paper: scaling problem, U tube experiments, hot leg experiments, high pressure tests

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

Sensitivity of the probability of failure to probability of detection curve regions

International Nuclear Information System (INIS)

Garza, J.; Millwater, H.

2016-01-01

Non-destructive inspection (NDI) techniques have been shown to play a vital role in fracture control plans, structural health monitoring, and ensuring availability and reliability of piping, pressure vessels, mechanical and aerospace equipment. Probabilistic fatigue simulations are often used in order to determine the efficacy of an inspection procedure with the NDI method modeled as a probability of detection (POD) curve. These simulations can be used to determine the most advantageous NDI method for a given application. As an aid to this process, a first order sensitivity method of the probability-of-failure (POF) with respect to regions of the POD curve (lower tail, middle region, right tail) is developed and presented here. The sensitivity method computes the partial derivative of the POF with respect to a change in each region of a POD or multiple POD curves. The sensitivities are computed at no cost by reusing the samples from an existing Monte Carlo (MC) analysis. A numerical example is presented considering single and multiple inspections. - Highlights: • Sensitivities of probability-of-failure to a region of probability-of-detection curve. • The sensitivities are computed with negligible cost. • Sensitivities identify the important region of a POD curve. • Sensitivities can be used as a guide to selecting the optimal POD curve.

Structural and piping issues in the design certification of advanced reactors

International Nuclear Information System (INIS)

Ali, S.A.; Terao, D.; Bagchi, G.

1996-01-01

The purpose of this paper is to discuss the design certification of structures and piping for evolutionary and passive advanced light water reactors. Advanced reactor designs are based on a set of assumed site-related parameters that are selected to envelop a majority of potential nuclear power plant sites. Multiple time histories are used as the seismic design basis in order to cover the majority of potential sites in the US. Additionally, design are established to ensure that surface motions at a particular site will not exceed the enveloped standard design surface motions. State-of-the-art soil-structure interaction (SSI) analyses have been performed for the advanced reactors, which include structure-to-structure interaction for all seismic Category 1 structures. Advanced technology has been utilized to exclude the dynamic effects of pipe rupture from structural design by demonstrating that the probability of pipe rupture is extremely low. For piping design, the advanced reactor vendors have developed design acceptance criteria (DAC) which provides the piping design analysis methods, design procedures, and acceptance criteria. In SECY-93-087 the NRC staff recommended that the Commission approve the approach to eliminate the OBE from the design of structures and piping in advanced reactors and provided guidance which identifies the necessary changes to existing seismic design criteria. The supplemental criteria address fatigue, seismic anchor motion, and piping stress limits when the OBE is eliminated

Analysis of two-phase flow induced vibrations in perpendiculary supported U-type piping systems

International Nuclear Information System (INIS)

Hiramatsu, Tsutomu; Komura, Yoshiaki; Ito, Atsushi.

1984-01-01

The perpose of this analysis is to predict the vibration level of a pipe conveying a two-phase flowing fluid. Experiments were carried out with a perpendiculary supported U-type piping system, conveying an air-water two-phase flow in a steady state condition. Fluctuation signals are observed by a void signal sensor, and power spectral densities and probability density functions are obtained from the void signals. Theoretical studies using FEM and an estimation of the exciting forces from the PSD of void signals, provided a good predictional estimation of vibration responses of the piping system. (author)

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

HPFRCC - Extruded Pipes

DEFF Research Database (Denmark)

Stang, Henrik; Pedersen, Carsten

1996-01-01

The present paper gives an overview of the research onHigh Performance Fiber Reinforced Cementitious Composite -- HPFRCC --pipes recently carried out at Department of Structural Engineering, Technical University of Denmark. The project combines material development, processing technique development......-w$ relationship is presented. Structural development involved definition of a new type of semi-flexiblecement based pipe, i.e. a cement based pipe characterized by the fact that the soil-pipe interaction related to pipe deformation is an importantcontribution to the in-situ load carrying capacity of the pipe...

Probabilistic procedure to evaluate integrity of degraded pipes under internal pressure and bending moment

International Nuclear Information System (INIS)

Roos, E.; Herter, K.-H.; Julisch, P.; Otremba, F.; Schuler, X.

2003-01-01

The determination of critical crack sizes or permissible/allowable loading levels in pipes with degraded pipe sections (circumferential cracks) for the assurance of component integrity is usually based on deterministic approaches. Therefore along with numerical calculational methods (finite element (FE) analyses) limit load calculations, such as e.g. the 'Plastic limit load concept' and the 'Flow stress concept' as well as fracture mechanics approximation methods as e.g. the R-curve method or the 'Ductile fracture handbook' and the R6-Method are currently used for practical application. Numerous experimental tests on both ferritic and austenitic pipes with different pipe dimensions were investigated at MPA Stuttgart. The geometries of the pipes were comparable to actual piping systems in Nuclear Power Plants, both BWR as well as PWR. Through wall cracks and part wall through cracks on the inside surface of the pipes were considered. The results of these tests were used to determine the flow stresses used within the limit load calculations. Therefore the deterministic concepts assessing the integrity of degraded pipes are available A new post-calculation of the above mentioned tests was performed using probabilistic approaches to assure the component integrity of degraded piping systems. As a result the calculated probability of failure was compared to experimental behaviour during the pipe test. Different reliability techniques were used for the verification of the probabilistic approaches. (author)

Pressure management strategies for water loss reduction in large-scale water piping networks: a review

CSIR Research Space (South Africa)

Adedeji, K

2017-06-01

Full Text Available . Therefore, in a WDN with kth number of pipes, the probability of a pipe breakage in the network as a result of the system pressure variations is estimated as [17] 4.35 3.738 5 8 0.0021 21.4 Pr 1 10 k kD D k k k e D e D     (5) where Prk...

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

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.

Experimental study on the thermostable property of aramid fiber reinforced PE-RT pipes

Directory of Open Access Journals (Sweden)

Guoquan Qi

2015-11-01

Full Text Available Flexible composite pipes are advantageous in ultra high strength, high modulus, pH and corrosion resistance and light weight, but there are still some hidden safety troubles because they are poorer in thermostable capacity. Therefore, test samples of flexible composite pipes were prepared with high-temperature polythene (PE-RT as the neck bush and aramid fiber as the reinforcement layer. Experimental study was conducted by using HPHT vessel and differential thermal scanner for different working conditions, different temperatures, whole-pipe pressure-bearing capacity and 1000 h viability. It is shown by the environmental compatibility test that high temperature has little effect on the weight, Vicat softening temperature, mechanical properties and structures of neck bush PE-RT, but exerts an obvious effect on the tensility and whole-pipe water pressure blasting of the reinforcement aramid fiber. Besides, the drop of whole-pipe pressure-bearing capacity is caused by deformation and breaking of aramid fibers when the reinforcement layer is under the force of internal pressure. Finally, disorientation and crystallization of molecular thermal motion occur with the rise of temperature, so amorphous orientation reduces, crystallinity factor and crystalline orientation factor increase gradually, thus, disorientation of macromolecular chains increases and tensile strength decreases. It is concluded that this type of flexible composite pipe can smoothly pass 1000 h viability test. And it is recommended that it be used in the situations with temperature not higher than 95 °C and internal pressure not higher than 4 MPa.

Counter-current flow limitation at hot leg pipe during reflux condensation cooling after small-break LOCA

International Nuclear Information System (INIS)

Jeong, Hae Yong; Ha, Sang Jun; Jo, Yung Jo; Jun, Hwang Yong

1999-01-01

The possibility of hot leg flooding is evaluated in case of a small-break loss-of-coolant accident in Korean Next Generation Reactor (KNGR) operating at the core power of 3983 MW normally. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The calculated results are compared with the existing flooding correlations. It is predicted that the hot leg flooding is excluded when two steam generators are available. It is also shown that the possibility of hot leg flooding under the operation with one steam generator is very low. Therefore, it can be said that the occurrence of hot leg flooding is unexpected when the reflux condensation cooling is maintained in steam generator tubes

Comparison of turbulent flow through hexagram and hexagon orifices in circular pipes using large-eddy simulation

Energy Technology Data Exchange (ETDEWEB)

Wang, Wei; Nicolleau, Franck C G A; Qin, Ning, E-mail: n.qin@sheffield.ac.uk [Department of Mechanical Engineering, The University of Sheffield, Sheffield, S1 3JD (United Kingdom)

2016-04-15

Characteristics of turbulent flow through a circular, a hexagon and a hexagram orifice with the same flow area in circular pipes are investigated using wall-modelled large-eddy simulation. Good agreements to available experimental data were obtained in both the mean velocity and turbulent kinetic energy. The hexagram orifice with alternating convex and concave corners introduces outwards radial velocity around the concave corners downstream of the orifice plate stronger than the hexagon orifice. The stronger outwards radial velocity transfers high momentum from the pipe centre towards the pipe wall to energize the orifice-forced vortex sheet rolling-up and leads to a delayed vortex break-down. Correspondingly, the hexagram has a more gradual flow recovery to a pipe flow and a reduced pressure drop than the hexagon orifice. Both the hexagon and hexagram orifices show an axis-switching phenomenon, which is observed from both the streamwise velocity and turbulent kinetic energy contours. To the best knowledge of the authors, this is the first comparison of orifice-forced turbulence development, mixing and flow dynamics between a regular and a fractal-based polygonal orifice. (paper)

ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-06-01

, mixture level, temperatur kelongsong, small break LOCA, RELAP5.   ABSTRACT ANALYSIS ON THE CORE CONDITION OF AP1000 ADVANCED POWER REACTOR DURING SMALL BREAK LOCA. Accident due to the loss of coolant from the reactor boundary is an anticipated design basis event in the design of power reactor adopting the Generation II up to IV technology. Small break LOCA leads to more significant impact on safety compared to the large break LOCA as shown in the Three-Mile Island (TMI. The focus of this paper is the small break LOCA analysis on the Generation III+ advanced power reactor of AP1000 by simulating three different initiating events, which are inadvertent opening of Automatic Depressurization System (ADS, double-ended break on one of Direct Vessel Injection (DVI pipe, and 10 inch diameter split break on one of cold leg pipe. Methodology used is by simulating the events on the AP1000 model developed using RELAP5/SCDAP/Mod3.4. The impact analyzed is the core condition during the small break LOCA consisting of the mixture level occurrence and the fuel cladding temperature transient. The results show that the mixture level for all small break LOCA events are above the active core height, which indicates no core uncovery event. The development of the mixture level affect the fuel cladding temperature transient, which shows a decreasingly trend after the break, and the effectifeness of core cooling. Those results are identical with the results of other code of NOTRUMP. The resulted core cooling is also due to the function of coolant injection from passive safety feature, which is unique in the AP1000 design. In overall, the result of analysis shows that the AP1000 model developed by the RELAP5 can be used for analysis of design basis accident considered in the AP1000 advanced power reactor. Keywords: analysis, mixture level, fuel cladding temperature, small break LOCA, RELAP5.

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

Improvement of C*-integral and Crack Opening Displacement Estimation Equations for Thin-walled Pipes with Circumferential Through-wall Cracks

International Nuclear Information System (INIS)

Park, Jeong Soon; Jhung, Myung Jo

2012-01-01

Since the LBB(Leak-Before-Break) concept has been widely applied to high energy piping systems in the pressurized water reactors, a number of engineering estimation methods had been developed for J-integral and COD values. However, those estimation methods were mostly reliable for relatively thick-walled pipes about R m /t=5 or 10. As the LBB concept might be considered in the design stage of the SFR (Sodium-cooled Fast Reactor) which has relatively thin-walled pipes due to its low design pressure, the applicability of current estimation methods should be investigated for thin-walled pipes. Along with the J-integral and COD, the estimation method for creep fracture mechanics parameters, C*- integral and COD rate, is required because operating temperature of SFR is high enough to induce creep in the structural materials. In this study, the applicability of the current C*- integral and COD estimation methods to thin-walled pipes is studied for a circumferential through-wall crack using the finite element (FE) method. Based on the FE results, enhancement of the current estimation methods is made

The Canadian approach to protection against postulated primary heat transport piping failures

International Nuclear Information System (INIS)

Jarman, B.L.

1985-10-01

In Canada, the Atomic Energy Control Act and Regulations stipulate in broad terms the requirements to be met by licensees. In addition, AECB staff have prepared licensing guides to amplify those requirements. For nuclear reactors, these include providing suitable protection against the consequences of failure of any pipe in the reactor cooling system. The suggested means of limiting the damage caused by whipping pipes or steam jets is by separation of equipment, installing barriers, or restraining piping. If, however, the designer can demonstrate that restraints are impractical or detrimental to safety, AECB staff may consider alternate arguments based on a demonstration that piping is likely to crack and then leak for a long time prior to rupture. This alternative approach would not be considered for ruptures having a high probability of defeating containment, damaging essential safety systems, or of disrupting flow to the core to the extent that fuel cooling could not be maintained

Suggestion for a homogenizer installation in LOFT small break two-phase measurement

International Nuclear Information System (INIS)

Rieger, G.

1981-07-01

The purpose of this task, which was performed as an Austrian inkind contribution for the INEL research program is a) the evaluation of literature concerning homogenizers to improve two phase flow measurements for the LOFT small break test series, b) design of a homogenizer and c) recommandation of the location of a homogenizer in the LOFT piping system. To optimize the location of the homogenizer LTSF-tests should be performed according to the suggestions in this paper. (author)

Probability via expectation

CERN Document Server

Whittle, Peter

1992-01-01

This book is a complete revision of the earlier work Probability which ap­ peared in 1970. While revised so radically and incorporating so much new material as to amount to a new text, it preserves both the aim and the approach of the original. That aim was stated as the provision of a 'first text in probability, de­ manding a reasonable but not extensive knowledge of mathematics, and taking the reader to what one might describe as a good intermediate level'. In doing so it attempted to break away from stereotyped applications, and consider applications of a more novel and significant character. The particular novelty of the approach was that expectation was taken as the prime concept, and the concept of expectation axiomatized rather than that of a probability measure. In the preface to the original text of 1970 (reproduced below, together with that to the Russian edition of 1982) I listed what I saw as the advantages of the approach in as unlaboured a fashion as I could. I also took the view that the text...

Development of gamma-ray densitometer and measurement of void fraction in instantaneous pipe rupture under BWR LOCA condition

International Nuclear Information System (INIS)

Yano, Toshikazu

1983-11-01

In order to clarify the transient mass flow rate under the instantaneous pipe rupture condition, it is necessary to use a highly sensitive void meter. Therefore, a high-response gamma-ray densitometer was developed for the measurement of void fraction variation caused by flashing vaporization of the high-pressure and -temperature water under the instantaneous pipe rupture accident. The measurement of void fraction was performed in the pipe rupture test under the BWR LOCA condition with a 6-inch diameter pipe. Initial conditions of the water were 6.86 MPa in pressure and the saturation temperature. To prove the reliability and accuracy, a calibration test by falling acrylic void simulators and an air injection test into cold water filled in the pipe were also conducted. The following results are obtained in the pipe rupture test. (1) The cone slit method is very useful to increase the measuring accuracy. (2) It is clearly observed that the apparent increase of void fraction occurs after the rarefaction wave passes. (3) The first maximum of void fraction occurs with some delay time after break. The following minimum void fraction concurs with the maximum pressure in the pressure recovering phenomena and with the maximum blowdown thrust force. (author)

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

Pipe damping

International Nuclear Information System (INIS)

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

1984-01-01

A program has been developed to assess the available piping damping data, to generate additional data and conduct seperate effects tests, and to establish a plan for reporting and storing future test results into a data bank. This effort is providing some of the basis for developing higher allowable damping values for piping seismic analyses, which will potentially permit removal of a considerable number of piping supports, particularly snubbers. This in turn will lead to more flexible piping systems which will be less susceptible to thermal cracking, will be easier to maintain and inspect, as well as less costly

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)

Insulated pipe clamp design

International Nuclear Information System (INIS)

Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

1980-01-01

Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. The design considerations and methods along with the development tests are presented. Special considerations to guard against adverse cracking of the insulation material, to maintain the clamp-pipe stiffness desired during a seismic event, to minimize clamp restraint on the pipe during normal pipe heatup, and to resist clamp rotation or spinning on the pipe are emphasized

Integrated failure probability estimation based on structural integrity analysis and failure data: Natural gas pipeline case

International Nuclear Information System (INIS)

Dundulis, Gintautas; Žutautaitė, Inga; Janulionis, Remigijus; Ušpuras, Eugenijus; Rimkevičius, Sigitas; Eid, Mohamed

2016-01-01

In this paper, the authors present an approach as an overall framework for the estimation of the failure probability of pipelines based on: the results of the deterministic-probabilistic structural integrity analysis (taking into account loads, material properties, geometry, boundary conditions, crack size, and defected zone thickness), the corrosion rate, the number of defects and failure data (involved into the model via application of Bayesian method). The proposed approach is applied to estimate the failure probability of a selected part of the Lithuanian natural gas transmission network. The presented approach for the estimation of integrated failure probability is a combination of several different analyses allowing us to obtain: the critical crack's length and depth, the failure probability of the defected zone thickness, dependency of the failure probability on the age of the natural gas transmission pipeline. A model's uncertainty analysis and uncertainty propagation analysis are performed, as well. - Highlights: • Degradation mechanisms of natural gas transmission pipelines. • Fracture mechanic analysis of the pipe with crack. • Stress evaluation of the pipe with critical crack. • Deterministic-probabilistic structural integrity analysis of gas pipeline. • Integrated estimation of pipeline failure probability by Bayesian method.

Piping failures in United States nuclear power plants 1961-1995

International Nuclear Information System (INIS)

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

1996-01-01

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

Mapping the double-strand breaks at the mating-type locus in fission yeast by genomic sequencing

DEFF Research Database (Denmark)

Nielsen, O; Egel, R; Nielsen, Olaf

1989-01-01

the ends are probably masked by tightly bound proteins and therefore the precise nature of the break could not be determined. Since the break is stable throughout the cell cycle, these proteins may function in vivo to confer structural stability on the chromosomes having the break. The implications...

Fundamentals of piping design

CERN Document Server

Smith, Peter

2013-01-01

Written for the piping engineer and designer in the field, this two-part series helps to fill a void in piping literature,since the Rip Weaver books of the '90s were taken out of print at the advent of the Computer Aid Design(CAD) era. Technology may have changed, however the fundamentals of piping rules still apply in the digitalrepresentation of process piping systems. The Fundamentals of Piping Design is an introduction to the designof piping systems, various processes and the layout of pipe work connecting the major items of equipment forthe new hire, the engineering student and the vetera

Heat pipe

International Nuclear Information System (INIS)

Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.

1986-01-01

A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)

An overview of erosion corrosion models and reliability assessment for corrosion defects in piping system

International Nuclear Information System (INIS)

Srividya, A.; Suresh, H.N.; Verma, A.K.; Gopika, V.; Santosh

2006-01-01

Piping systems are part of passive structural elements in power plants. The analysis of the piping systems and their quantification in terms of failure probability is of utmost importance. The piping systems may fail due to various degradation mechanisms like thermal fatigue, erosion-corrosion, stress corrosion cracking and vibration fatigue. On examination of previous results, erosion corrosion was more prevalent and wall thinning is a time dependent phenomenon. The paper is intended to consolidate the work done by various investigators on erosion corrosion in estimating the erosion corrosion rate and reliability predictions. A comparison of various erosion corrosion models is made. The reliability predictions based on remaining strength of corroded pipelines by wall thinning is also attempted. Variables in the limit state functions are modelled using normal distributions and Reliability assessment is carried out using some of the existing failure pressure models. A steady state corrosion rate is assumed to estimate the corrosion defect and First Order Reliability Method (FORM) is used to find the probability of failure associated with corrosion defects over time using the software for Component Reliability evaluation (COMREL). (author)

Laser fluorescent method for monitoring leaks from petrol pipes based on the neural network algorithm

Directory of Open Access Journals (Sweden)

M. L. Belov

2014-01-01

Full Text Available Current systems for monitoring leaks from petrol pipes can detect large leaks only, and their sensitivity limit is about 1% of the whole petrol pipe’s capacity. In this paper, a problem of remote detection of small leaks (less than 1% from petrol pipes was considered. One of possible variations of such a system is a monitoring system of oil pollution at the earth surface along the petrol pipe. In this paper experimentally obtained data such as fluorescence spectra of oil products (crude oil, light-end oil products, heavy oil products, various earth surfaces (soil, vegetation, water, asphalt and oil products spilled over various earth's surface were used for the excitation wavelength of 266 nm. It was shown that use of the laser method based on detection of fluorescence radiation within three narrow spectral bands and a neural network algorithm of measured data processing allowed one to detect oil pollution on the earth surface with a probability of correct classification close to 1 and low probability of false alarm.

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

Structural dynamics and fracture mechanics calculations of the behaviour of a DN 425 test piping system subjected to transient loading by water hammer

International Nuclear Information System (INIS)

Kussmaul, K.; Kobes, E.; Diem, H.; Schrammel, D.; Brosi, S.

1994-01-01

Within the scope of the German HDR safety programme, several tests were carried out to investigate transient pipe loading initiated by a simulated double-ended guillotine break event, and subsequent closure of a feedwater check valve (water hammer, blow-down). Numerical analyses by means of finite element programmes were performed in parallel to the experiments. Using water hammer tests of a DN 425 piping system with predamaged components, the procedure of such analyses will be demonstrated. The results are presented, beginning with structural dynamic calculations of the undamaged piping; followed by coupling of structural dynamics and fracture mechanics computations with simple flaw elements (line spring); and finishing with costly three-dimensional fracture mechanics analyses. A good description of the real piping behaviour can be made by the numerical methods, even in the case of high plastification processes. ((orig.))

Heat pipe and method of production of a heat pipe

International Nuclear Information System (INIS)

Kemp, R.S.

1975-01-01

The heat pipe consists of a copper pipe in which a capillary network or wick of heat-conducting material is arranged in direct contact with the pipe along its whole length. Furthermore, the interior space of the tube contains an evaporable liquid for pipe transfer. If water is used, the capillary network consists of, e.g., a phosphorus band network. To avoid contamination of the interior of the heat pipe during sealing, its ends are closed by mechanical deformation so that an arched or plane surface is obtained which is in direct contact with the network. After evacuation of the interior space, the remaining opening is closed with a tapered pin. The ratio wall thickness/tube diameter is between 0.01 and 0.6. (TK/AK) [de

Pipe drafting and design

CERN Document Server

Parisher, Roy A; Parisher

2000-01-01

Pipe designers and drafters provide thousands of piping drawings used in the layout of industrial and other facilities. The layouts must comply with safety codes, government standards, client specifications, budget, and start-up date. Pipe Drafting and Design, Second Edition provides step-by-step instructions to walk pipe designers and drafters and students in Engineering Design Graphics and Engineering Technology through the creation of piping arrangement and isometric drawings using symbols for fittings, flanges, valves, and mechanical equipment. The book is appropriate primarily for pipe

Study of check valve slamming in a BWR feedwater system following a postulated pipe break

International Nuclear Information System (INIS)

Safwat, H.H.; Arastu, A.H.; Norman, A.

1985-01-01

This study deals with a swing check valve slamming due to a break at relatively short distance from the valve. Under this situation, substantial flashing occurs near the valve and the result of the study are subject to what is believed to be a conservative simplifying assumption, i.e., the hydrodynamic moment acting on the valve during the transient is represented by resultant moment due to the pressure differential across the valve. It is believed that vapor voids forming at the valve would actually reduce the disk impact velocities in comparison to those predicted under this simplifying assumption. A technique used to represent a double-ended break through hypothetical valves may have some influence on the results particularly for long break opening times. The study has yielded good insight to help understand the complex problem. The study has focused on some parameters and the reader may raise questions on the effects of other parameters. Nevertheless, the present study underlines the complexity facing analysts dealing with this transient using analytical methods. Though some experimental data are available, the authors believe that an experimental study (recognizing the complexity of the experimental setup and instrumentation), would be quite useful. It can provide answers to questions facing analysts dealing with this problem and thus avoid unnecessary conservatisms due to uncertainties in input data

Provisions and active measures to preclude guillotine breaks in piping systems of pressurized water reactors

International Nuclear Information System (INIS)

Dorner, H.; Michel, E.

1983-01-01

The conditions and active measures which preclude a spontaneous failure of pipings are shown. With the basic safety concept a quality standard is achieved characterized by high-grade material properties, a structure that is adequate to the loads to which the components will be subjected in service and is amenable to inspection, precise load and stress evaluation, optimized manufacturing and operation monitoring. The possible failure types are described and the safety against failure is assessed. (author) [pt

Methods of assessing the leak-before-break behaviour of pressurized components

International Nuclear Information System (INIS)

Goerner, F.

1984-01-01

A general overview of the parameters is first given, which are important for the stress and service life of a pressurized component. The individual parameters are discussed, where the main points are the calculation of stress intensity factors, the fatigue behaviour and the calculation of plastic limiting loads and elastic-plastic failure factors (COD and J integral), using the Dugdale model. In a final chapter, the leak-before-break diagrams are given and compared for different methods of calculation for pipes with longitudinal and circumferential cracks and for flat plates. (orig./HP) [de

Genetics of x-ray induced double strand break repair in saccharomyces cerevisiae

International Nuclear Information System (INIS)

Budd, M.E.

1982-07-01

The possible fates of x-ray-induced double-strand breaks in Saccharomyces cerevisiae were examined. One possible pathway which breaks can follow, the repair pathway, was studied by assaying strains with mutations in the RAD51, RAD54, and RAD57 loci for double-strand break repair. In order of increasing radiation sensitivity one finds: rad57-1(23 0 )> rad51-1(30 0 )> rad54-3(36 0 ). At 36 0 , rad54-3 cells cannot repair double-strand breaks, while 23 0 , they can. Strains with the rad57-1 mutation can rejoin broken chromosomes at both temperatures. However, the low survival at 36 0 shows that the assay is not distinguishing large DNA fragments which allow cell survival from those which cause cell death. A rad51-1 strain could also rejoin broken chromosomes, and was thus capable of incomplete repair. The data can be explained with the hypothesis that rad54-3 cells are blocked in an early step of repair, while rad51-1 and rad57-1 strains are blocked in a later step of repair. The fate of double-strand breaks when they are left unrepaired was investigated with the rad54-3 mutation. If breaks are prevented from entering the RAD54 repair pathway they become uncommitted lesions. These lesions are repaired slower than the original breaks. One possible fate for an uncommitted lesion is conversion into a fixed lesion, which is likely to be an unrepairable or misrepaired double-strand break. The presence of protein synthesis after irradiation increases the probability that a break will enter the repair pathway. Evidence shows that increased probability of repair results from enhanced synthesis of repair proteins shortly after radiation

Leak before break application in French PWR plants under operation

Energy Technology Data Exchange (ETDEWEB)

Faidy, C. [EDF SEPTEN, Villeurbanne (France)

1997-04-01

Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

Non-destructive qualification tests for ITER cryogenic axial insulating breaks

International Nuclear Information System (INIS)

Kosek, Jacek; Lopez, Roberto; Tommasini, Davide; Rodriguez-Mateos, Felix

2014-01-01

In the ITER superconducting magnets the dielectric separation between the CICC (Cable-In-Conduit Conductors) and the helium supply pipes is made through the so-called insulating breaks (IB). These devices shall provide the required dielectric insulation at a 30 kV level under different types of stresses and constraints: thermal, mechanical, dielectric and ionizing radiations. As part of the R and D program, the ITER Organization launched contracts with industrial companies aimed at the qualification of the manufacturing techniques. After reviewing the main functional aspects, this paper describes and discusses the protocol established for non-destructive qualification tests of the prototypes

Non-destructive qualification tests for ITER cryogenic axial insulating breaks

Energy Technology Data Exchange (ETDEWEB)

Kosek, Jacek [Wroclaw University of Technology, Wybrzeze Wyspianskiego 27, 50-370 Wroclaw, Poland and CERN, Geneva 23,CH-1211 (Switzerland); Lopez, Roberto; Tommasini, Davide [CERN, Geneva 23,CH-1211 (Switzerland); Rodriguez-Mateos, Felix [CERN, Geneva 23,CH-1211, Switzerland and ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul lez Durance (France)

2014-01-29

In the ITER superconducting magnets the dielectric separation between the CICC (Cable-In-Conduit Conductors) and the helium supply pipes is made through the so-called insulating breaks (IB). These devices shall provide the required dielectric insulation at a 30 kV level under different types of stresses and constraints: thermal, mechanical, dielectric and ionizing radiations. As part of the R and D program, the ITER Organization launched contracts with industrial companies aimed at the qualification of the manufacturing techniques. After reviewing the main functional aspects, this paper describes and discusses the protocol established for non-destructive qualification tests of the prototypes.

DENSE CORES IN THE PIPE NEBULA: AN IMPROVED CORE MASS FUNCTION

International Nuclear Information System (INIS)

Rathborne, J. M.; Lada, C. J.; Muench, A. A.; Alves, J. F.; Kainulainen, J.; Lombardi, M.

2009-01-01

In this paper, we derive an improved core mass function (CMF) for the Pipe Nebula from a detailed comparison between measurements of visual extinction and molecular-line emission. We have compiled a refined sample of 201 dense cores toward the Pipe Nebula using a two-dimensional threshold identification algorithm informed by recent simulations of dense core populations. Measurements of radial velocities using complimentary C 18 O (1-0) observations enable us to cull out from this sample those 43 extinction peaks that are either not associated with dense gas or are not physically associated with the Pipe Nebula. Moreover, we use the derived C 18 O central velocities to differentiate between single cores with internal structure and blends of two or more physically distinct cores, superposed along the same line of sight. We then are able to produce a more robust dense core sample for future follow-up studies and a more reliable CMF than was possible previously. We confirm earlier indications that the CMF for the Pipe Nebula departs from a single power-law-like form with a break or knee at M ∼ 2.7 ± 1.3 M sun . Moreover, we also confirm that the CMF exhibits a similar shape to the stellar initial mass function (IMF), but is scaled to higher masses by a factor of ∼4.5. We interpret this difference in scaling to be a measure of the star formation efficiency (22% ± 8%). This supports earlier suggestions that the stellar IMF may originate more or less directly from the CMF.

Heat pipes and use of heat pipes in furnace exhaust

Science.gov (United States)

Polcyn, Adam D.

2010-12-28

An array of a plurality of heat pipe are mounted in spaced relationship to one another with the hot end of the heat pipes in a heated environment, e.g. the exhaust flue of a furnace, and the cold end outside the furnace. Heat conversion equipment is connected to the cold end of the heat pipes.

Drill pipe bridge plug

International Nuclear Information System (INIS)

Winslow, D.W.; Brisco, D.P.

1991-01-01

This patent describes a method of stopping flow of fluid up through a pipe bore of a pipe string in a well. It comprises: lowering a bridge plug apparatus on a work string into the pipe string to a position where the pipe bore is to be closed; communicating the pipe bore below a packer of the bridge plug apparatus through the bridge plug apparatus with a low pressure zone above the packer to permit the fluid to flow up through the bridge plug apparatus; engaging the bridge plug apparatus with an internal upset of the pipe string; while the fluid is flowing up through the bridge plug apparatus, pulling upward on the work string and the bridge plug apparatus and thereby sealing the packer against the pipe bore; isolating the pipe bore below the packer from the low pressure zone above the packer and thereby stopping flow of the fluid up through the pipe bore; disconnecting the work string from the bridge plug apparatus; and maintaining the bridge plug apparatus in engagement with the internal upset and sealed against the pipe bore due to an upward pressure differential applied to the bridge plug apparatus by the fluid contained therebelow

TACIS 91: Application of leak-before-break concept in VVER 440-230

Energy Technology Data Exchange (ETDEWEB)

Bartholome, G.; Faidy, C.; Franco, C. [and others

1997-04-01

The applicability of the leak-before-break (LBB) concept for primary piping in the first generation of WWER type plants in Russia is investigated. The procedures for LBB behavior used in France and Germany are applied, and the evaluation is discussed within the framework of the European Technical Assistance for the Community of Independent States (TACIS) project. Emphasis is placed on experimental validation of national and international engineering practice for evaluating and optimizing existing installations. Design criteria of WWER plants are compared to western standard design.

Insulated pipe clamp design

International Nuclear Information System (INIS)

Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

1980-01-01

Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. 5 refs

Miniature Heat Pipes

Science.gov (United States)

1997-01-01

Small Business Innovation Research contracts from Goddard Space Flight Center to Thermacore Inc. have fostered the company work on devices tagged "heat pipes" for space application. To control the extreme temperature ranges in space, heat pipes are important to spacecraft. The problem was to maintain an 8-watt central processing unit (CPU) at less than 90 C in a notebook computer using no power, with very little space available and without using forced convection. Thermacore's answer was in the design of a powder metal wick that transfers CPU heat from a tightly confined spot to an area near available air flow. The heat pipe technology permits a notebook computer to be operated in any position without loss of performance. Miniature heat pipe technology has successfully been applied, such as in Pentium Processor notebook computers. The company expects its heat pipes to accommodate desktop computers as well. Cellular phones, camcorders, and other hand-held electronics are forsible applications for heat pipes.

Influence of Crack Morphology on Leak Before Break Margins

International Nuclear Information System (INIS)

Weilin Zang

2007-11-01

The purpose of the project is to evaluate the deterministic LBB-margins for different pipe systems in a Swedish PWR-plant and using different crack morphology parameters. Results: - The influence of crack morphology on Leak Before Break (LBB) margins is studied. The subject of the report is a number of LBB-submittals to SKI where deterministic LBB-margins are reported. These submittals typically uses a surface roughness of 0.0762 mm (300 microinch) and number of turns equal to zero and an in-house code for the leak rate evaluations. The present report has shown that these conditions give the largest LBB-margins both in terms of the quotient between the critical crack length and the leakage crack size and for the leak rate margin. - Crack morphology parameters have a strong influence on the leak rate evaluations. Using the SQUIRT code and more recent recommendations for crack morphology parameters, it is shown that in many cases the evaluated margins, using 1 gpm as the reference leak rate detection limit, are below the safety factor of 2 on crack size and 10 on leak rate, which is generally required for LBB approval. - The effect of including weld residual stresses on the LBB margins is also investigated. It is shown that for the two examples studied, weld residual stresses were important for the small diameter thin wall pipe whereas it was negligible for the large diameter thick wall pipe which had a self-balanced weld residual stress distribution

Riser pipe elevator

Energy Technology Data Exchange (ETDEWEB)

Bennett, W.; Jimenez, A.F.

1987-09-08

This patent describes a method for storing and retrieving a riser pipe, comprising the steps of: providing an upright annular magazine comprised of an inside annular wall and an outside annular wall, the magazine having an open top; storing the riser pipe in a substantially vertically oriented position within the annular magazine; and moving the riser pipe upwardly through the open top of the annular magazine at an angle to the vertical along at least a portion of the length of the riser pipe.

The J-resistance curve Leak-before-Break test program on material for the Darlington Nuclear Generating Station

International Nuclear Information System (INIS)

Mukherjee, B.

1988-01-01

The Darlington Leak-Before-Break (DLBB) approach has been developed for large diameter (21, 22, 24 inch) SA106B heat transport (HT) piping and SA105 fittings as a design alternative to pipewhip restraints and in recognition of the questionable benefits of providing such restraints. Ontario Hydro's DLBB approach is based on the elastic plastic fracture mechanics method. In this test program, J-resistance curves were determined from actual pipe heats that were used in the construction of the Darlington heat transport systems (Units 1 and 2). Test blocks were prepared using four different welding procedures for nuclear Class I piping. The test program was designed to take into account the effect of various factors such as test temperature, crack plane orientation, welding effects, etc., which have influence on fracture properties. A total of 91 tests were conducted. An acceptable lower bound J-resistance curve for the piping steels was obtained by machining maximum thickness specimens from the pipes and by testing side grooved compact tension specimens. Test results showed that all pipes, welds and heat-affected zone materials within the scope of the DLBB program exhibited uppershelf toughness behaviour. All specimens showed high crack initiation toughness Jsub(lc), rising J-resistance curve and stable and ductile crack extension. Toughness of product forms depended on the direction of crack extension (circumferential versus axial crack orientation). Toughness of DLBB welds and parent materials at 250 0 C was lower than that at 20 0 C. (author)

Dam Break Analysis of Embankment Dams Considering Breach Characteristics

Directory of Open Access Journals (Sweden)

Abolfazl Shamsaei

2004-05-01

Full Text Available The study of dam's break, needs the definition of various parameters such as the break cause, its type, its dimension and the duration of breach development. The precise forecast for different aspects of the breach is one of the most important factors for analyzing it in embankment dam. The characteristics of the breach and determination of their vulnerability has the most effect on the waves resulting from dam break. Investigating, about the parameters of the breach in "Silveh" earth dam have been determined using the suitable model. In Silve dam a trapezoid breach with side slope z=0.01m and the average base line b=80m was computed. The duration of the breaches development is 1.9 hour. Regarding the above results and the application of DAM Break software the consequences of the probable break of the dam was determined. The analysis of the results of water covering of the city of Piranshahr located 12km from silve dam confirms that in 3 hours the water will reach the height (level of 1425 meters.

PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping

International Nuclear Information System (INIS)

Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.

1975-01-01

To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report

Fragmentation in DNA double-strand breaks

International Nuclear Information System (INIS)

Wei Zhiyong; Suzhou Univ., Suzhou; Zhang Lihui; Li Ming; Fan Wo; Xu Yujie

2005-01-01

DNA double strand breaks are important lesions induced by irradiations. Random breakage model or quantification supported by this concept is suitable to analyze DNA double strand break data induced by low LET radiation, but deviation from random breakage model is more evident in high LET radiation data analysis. In this work we develop a new method, statistical fragmentation model, to analyze the fragmentation process of DNA double strand breaks. After charged particles enter the biological cell, they produce ionizations along their tracks, and transfer their energies to the cells and break the cellular DNA strands into fragments. The probable distribution of the fragments is obtained under the condition in which the entropy is maximum. Under the approximation E≅E 0 + E 1 l + E 2 l 2 , the distribution functions are obtained as exp(αl + βl 2 ). There are two components, the one proportional to exp(βl 2 ), mainly contributes to the low mass fragment yields, the other component, proportional to exp(αl), decreases slowly as the mass of the fragments increases. Numerical solution of the constraint equations provides parameters α and β. Experimental data, especially when the energy deposition is higher, support the statistical fragmentation model. (authors)

Piping research program plan

International Nuclear Information System (INIS)

1988-09-01

This document presents the piping research program plan for the Structural and Seismic Engineering Branch and the Materials Engineering Branch of the Division of Engineering, Office of Nuclear Regulatory Research. The plan describes the research to be performed in the areas of piping design criteria, environmentally assisted cracking, pipe fracture, and leak detection and leak rate estimation. The piping research program addresses the regulatory issues regarding piping design and piping integrity facing the NRC today and in the foreseeable future. The plan discusses the regulatory issues and needs for the research, the objectives, key aspects, and schedule for each research project, or group of projects focussing of a specific topic, and, finally, the integration of the research areas into the regulatory process is described. The plan presents a snap-shot of the piping research program as it exists today. However, the program plan will change as the regulatory issues and needs change. Consequently, this document will be revised on a bi-annual basis to reflect the changes in the piping research program. (author)

The probability of an encounter of two Brownian particles before escape

International Nuclear Information System (INIS)

Holcman, D; Kupka, I

2009-01-01

We study the probability of meeting of two Brownian particles before one of them exits a finite interval. We obtain an explicit expression for the probability as a function of the initial distance between the two particles using the Weierstrass elliptic function. We also find the law of the meeting location. Brownian simulations show the accuracy of our analysis. Finally, we discuss some applications to the probability that a double-strand DNA break repairs in confined environments.

Analytical studies of blowdown thrust force and dynamic response of pipe at pipe rupture accident

International Nuclear Information System (INIS)

Miyazaki, Noriyuki

1985-01-01

The motion of a pipe due to blowdown thrust when the pipe broke is called pipe whip. In LWR power plants, by installing restraints, the motion of a pipe when it broke is suppressed, so that the damage does not spread to neighboring equipment by pipe whip. When the pipe whip of a piping system in a LWR power plant is analyzed, blowdown thrust and the dynamic response of a pipe-restraint system are calculated with a computer. The blowdown thrust can be calculated by using such physical quantities as the pressure, flow velocity, density and so on in the system at the time of blowdown, obtained by the thermal-fluid analysis code at LOCA. The dynamic response of a piping-restraint system can be determined by the stress analysis code using finite element method taking the blowdown thrust as an external force acting on the piping. In this study, the validity of the analysis techniques was verified by comparing with the experimental results of the measurement of blowdown thrust and the pipe whip of a piping-restraint system, carried out in the Japan Atomic Energy Research Institute. Also the simplified analysis method to give the maximum strain on a pipe surface is presented. (Kako, I.)

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

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

Solar heating pipe

Energy Technology Data Exchange (ETDEWEB)

Hinson-Rider, G.

1977-10-04

A fluid carrying pipe is described having an integral transparent portion formed into a longitudinally extending cylindrical lens that focuses solar heat rays to a focal axis within the volume of the pipe. The pipe on the side opposite the lens has a heat ray absorbent coating for absorbing heat from light rays that pass through the focal axis.

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

Large-bore pipe decontamination

International Nuclear Information System (INIS)

Ebadian, M.A.

1998-01-01

The decontamination and decommissioning (D and D) of 1200 buildings within the US Department of Energy-Office of Environmental Management (DOE-EM) Complex will require the disposition of miles of pipe. The disposition of large-bore pipe, in particular, presents difficulties in the area of decontamination and characterization. The pipe is potentially contaminated internally as well as externally. This situation requires a system capable of decontaminating and characterizing both the inside and outside of the pipe. Current decontamination and characterization systems are not designed for application to this geometry, making the direct disposal of piping systems necessary in many cases. The pipe often creates voids in the disposal cell, which requires the pipe to be cut in half or filled with a grout material. These methods are labor intensive and costly to perform on large volumes of pipe. Direct disposal does not take advantage of recycling, which could provide monetary dividends. To facilitate the decontamination and characterization of large-bore piping and thereby reduce the volume of piping required for disposal, a detailed analysis will be conducted to document the pipe remediation problem set; determine potential technologies to solve this remediation problem set; design and laboratory test potential decontamination and characterization technologies; fabricate a prototype system; provide a cost-benefit analysis of the proposed system; and transfer the technology to industry. This report summarizes the activities performed during fiscal year 1997 and describes the planned activities for fiscal year 1998. Accomplishments for FY97 include the development of the applicable and relevant and appropriate regulations, the screening of decontamination and characterization technologies, and the selection and initial design of the decontamination system

Development of Pipe Holding Mechanism for Pipe Inspection Robot Using Flexible Pneumatic Cylinder

Directory of Open Access Journals (Sweden)

Choi Kyujun

2016-01-01

Full Text Available A pipe inspection robot is useful to reduce the inspection cost. In the previous study, a novel pipe inspection robot using a flexible pneumatic cylinder that can move forward along to the pipe by changing the robot’s body naturally was proposed and tested. In this paper, to improve its mobility for a corner of a pipe, the thin pipe holding mechanism using pneumatic bellows was proposed and tested. As a result of its driving test, the holding performance of the mechanism was confirmed.

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

Development of modified piping evaluation diagram for LBB application to Korean next generation reactor

International Nuclear Information System (INIS)

Huh, Nam Su; Kim, Young Jin; Pyo, Chang ryul; Yu, Young Jun; Yang, Jun Seog

1999-01-01

Recently, the Piping Evaluation Diagram (PED) is accepted in nuclear industry for simple application of Leak-Before-Break (LBB) concept to piping system. By utilizing the PED, the LBB concept is applied before the piping layout is finalized. However, the developed PED may have to be modified to account for the difference between the material properties of the PED development stage and those of the assembly stage. The objective of this paper is to develop the modified PED which can account for the variation of material properties. For this purpose, a parametric study was performed to investigate the effect of stress-strain curve on the detectable crack length and the effect of fracture resistance curve on the LBB allowable load. Finite element analyses were also performed to investigate the effect of stress-strain curve on the LBB allowable load. Finally a modified PED is developed as a function of crack length (DLC and 2xDLC) and the allowable Safe shutdown Earthquake (SSE) load. By adopting the modified PED, the variation of material properties can be considered in the LBB analysis and the computer runs required for the LBB analysis can be considerably reduced

The nature thickness pipe element testing method to validate the application of LBB conception

Energy Technology Data Exchange (ETDEWEB)

Vasilchenko, G.S.; Artemyev, V.I.; Merinov, G.N. [and others

1997-04-01

To validate the application of leak before break analysis to the VVER-1000 reactor, a procedure for testing a large-scale specimen on electrohydraulic machinery was developed. Steel pipe with a circular weld and stainless cladding inside was manufactured and large-scale longitudinal cross-sections were cut. The remaining parts of the weld after cut out were used to determination standard tensile mechanical properties, critical temperature of brittlness and for manufacture of compact specimens. Experimental mechanical properties of the weld are summarized.

The nature thickness pipe element testing method to validate the application of LBB conception

International Nuclear Information System (INIS)

Vasilchenko, G.S.; Artemyev, V.I.; Merinov, G.N.

1997-01-01

To validate the application of leak before break analysis to the VVER-1000 reactor, a procedure for testing a large-scale specimen on electrohydraulic machinery was developed. Steel pipe with a circular weld and stainless cladding inside was manufactured and large-scale longitudinal cross-sections were cut. The remaining parts of the weld after cut out were used to determination standard tensile mechanical properties, critical temperature of brittlness and for manufacture of compact specimens. Experimental mechanical properties of the weld are summarized

Application of Leak Before Break concept in 316LN austenitic steel pipes welded using 316L; Aplicação do conceito 1vazamento antes da falha' (Leak Before Break) em tubulações de aço 316LN soldado com metal de adição 316L

Energy Technology Data Exchange (ETDEWEB)

Cunto, Gabriel Giannini de

2017-07-01

This work presents a study of application of the Leak Before Break (LBB) concept, usually applied in nuclear power plants, in a pipe made from steel AISI type 316LN welded a coated electrode AISI type 316L. LBB concept is a criterion based on fracture mechanics analysis to show that a crack leak, present in a pipe, can be detected by leak detection systems, before this crack reaches a critical size that results in pipe fail. In the studied pipe, tensile tests and Ramberg-Osgood analyses were performed, as well as fracture toughness tests for obtaining the material resistance curve J-R. The tests were performed considering the base metal, weld and heat affected zone (HAZ), at the same operating temperatures of a nuclear power plant. For the mechanical properties found in these tests, load limit analyses were performed in order to determine the size of a crack which could cause a detectable leakage and the critical crack size, considering failure by plastic collapse. For the critical crack size found in the weld, which is the region that presented the lowest toughness, Integral J and tearing modulus T analyses were performed, considering failure by tearing instability. Results show a well-defined behavior between the base metal, HAZ and weld zones, where the base metal has a high toughness behavior, the weld has a low toughness behavior and the HAZ showed intermediate mechanical properties between the base metal and the weld. Using the PICEP software, the leak rate curves versus crack size and also the critical crack size were determined by considering load limit analysis. It was observed that after a certain crack size, the leak rate in base metal is much higher than for the HAZ and the weld, considering the same crack length. This occurs because in the base metal crack, it is expected that the crack grows in a more rounded form due to its higher toughness. The lowest critical crack size was found for the base metal presenting circumferential cracks. For the

On the ether-like Lorentz-breaking actions

International Nuclear Information System (INIS)

Petrov, A.Yu; Nascimento, J.R.; Gomes, M.; Silva, A. J. da

2011-01-01

We demonstrate the generation of the CPT-even, ether-like Lorentz-breaking actions for the scalar and electro-magnetic fields via their appropriate Lorentz-breaking coupling to spinor fields in three, four and five space-time dimensions. Besides, we show that the ether-like terms for the spinor field also can be generated as a consequence of the same couplings. The key result which will be presented here is the finiteness of the ether-like term for the electromagnetic field not only in three and five space-time dimensions where it is natural due to known effects of the dimensional regularization but also in four space-time dimensions. Moreover, we present the calculation of the last result within different calculational schemes and conclude that the result for the four-dimensional ether-like term for the electromagnetic field essentially depending on the calculation scheme, similarly to the result for the Carroll-Field-Jackiw (CFJ) term which probably signalizes a possibility for arising of a new anomaly. Also we discuss the dispersion relations in the theories with ether-like Lorentz-breaking terms which allows to discuss the consistency of the Lorentz-breaking modified theories for different (space-like or time-like) Lorentz-breaking vectors and find the tree-level effective (Breit) potential for fermion scattering and the one-loop effective potential corresponding to the action of the scalar field. (author)

Oscillating heat pipes

CERN Document Server

Ma, Hongbin

2015-01-01

This book presents the fundamental fluid flow and heat transfer principles occurring in oscillating heat pipes and also provides updated developments and recent innovations in research and applications of heat pipes. Starting with fundamental presentation of heat pipes, the focus is on oscillating motions and its heat transfer enhancement in a two-phase heat transfer system. The book covers thermodynamic analysis, interfacial phenomenon, thin film evaporation,  theoretical models of oscillating motion and heat transfer of single phase and two-phase flows, primary  factors affecting oscillating motions and heat transfer,  neutron imaging study of oscillating motions in an oscillating heat pipes, and nanofluid’s effect on the heat transfer performance in oscillating heat pipes.  The importance of thermally-excited oscillating motion combined with phase change heat transfer to a wide variety of applications is emphasized. This book is an essential resource and learning tool for senior undergraduate, gradua...

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

Heat pipes

CERN Document Server

Dunn, Peter D

1994-01-01

It is approximately 10 years since the Third Edition of Heat Pipes was published and the text is now established as the standard work on the subject. This new edition has been extensively updated, with revisions to most chapters. The introduction of new working fluids and extended life test data have been taken into account in chapter 3. A number of new types of heat pipes have become popular, and others have proved less effective. This is reflected in the contents of chapter 5. Heat pipes are employed in a wide range of applications, including electronics cooling, diecasting and injection mo

An in-pipe mobile micromachine using fluid power. A mechanism adaptable to pipe diameters

International Nuclear Information System (INIS)

Yoshida, Kazuhiro; Yokota, Shinichi; Takahashi, Ken

2000-01-01

To realize micro maintenance robots for small diameter pipes of nuclear reactors and so on, high power in-pipe mobile micromachines have been required. The authors have proposed the bellows microactuator using fluid power and have tried to apply the actuators to in-pipe mobile micromachines. In the previous papers, some inchworm mobile machine prototypes with 25 mm in diameter are fabricated and the traveling performances are experimentally investigated. In this paper, to miniaturize the in-pipe mobile machine and to make it adaptable to pipe diameters, firstly, a simple rubber-tube actuator constrained with a coil-spring is proposed and the static characteristics are investigated. Secondly, a supporting mechanism which utilizes a toggle mechanism and is adaptable to pipe diameters is proposed and the supporting forces are investigated. Finally, an in-pipe mobile micromachine for pipe with 4 - 5 mm in diameter is fabricated and the maximum traveling velocity of 7 mm/s in both ahead and astern movements is experimentally verified. (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.

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

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.

Study on pressure pulsation and piping vibration of complex piping of reciprocating compressor

International Nuclear Information System (INIS)

Xu Bin; Feng Quanke; Yu Xiaoling

2008-01-01

This paper presents a preliminary research on the piping vibration and pressure pulsation of reciprocating compressor piping system. On the basis of plane wave theory, the calculation of gas column natural frequency and pressure pulsation in complex pipelines is done by using the transfer matrix method and stiffness matrix method, respectively. With the discretization method of FEM, a mathematical model for calculating the piping vibration and stress of reciprocating compressor piping system is established, and proper boundary conditions are proposed. Then the structural modal and stress of the piping system are calculated with CAESAR II. The comparison of measured and calculated values found that the one dimensional wave equation can accurately calculate the natural frequency and pressure pulsation in gas column of piping system for reciprocating compressor. (authors)

Failure Analysis of PRDS Pipe in a Thermal Power Plant Boiler

Science.gov (United States)

Ghosh, Debashis; Ray, Subrata; Mandal, Jiten; Mandal, Nilrudra; Shukla, Awdhesh Kumar

2018-04-01

The pressure reducer desuperheater (PRDS) pipeline is used for reducing the pressure and desuperheating of the steam in different auxiliary pipeline. When the PRDS pipeline is failed, the reliability of the boiler is affected. This paper investigates the probable cause/causes of failure of the PRDS tapping line. In that context, visual inspection, outside diameter and wall thickness measurement, chemical analysis, metallographic examination and hardness measurement are conducted as part of the investigative studies. Apart from these tests, mechanical testing and fractographic analysis are also conducted as supplements. Finally, it has been concluded that the PRDS pipeline has mainly failed due to graphitization due to prolonged exposure of the pipe at higher temperature. The improper material used is mainly responsible for premature failure of the pipe.

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)

French regulatory practice and reflections on the leak-before-break concept

International Nuclear Information System (INIS)

Barrachin, B.; Jamet, P.

1986-01-01

For nuclear units in the design phase, the vendor (FRAMATOME) and the utility (EDF) examine the possibility of applying the leak-before-break criterion to a certain number of circuits, particularly the primary circuit and various parts of the secondary circuit. EDF has not completed its reflection on the leak-before-break approach and consequently, French safety authorities have not been officially notified of any request for assessment of the acceptability of such an approach. Nevertheless, from a strict safety point of view, it is possible to determine minimal conditions required in the implementation of such a course of action. Firstly, removing pipe breaks from the design basis of the plant will significantly weaken the application of the defense in-depth concept. This must be counter-balanced by reinforced manufacturing non destructive examinations and a stringent in-service inspection. Secondly, the demonstration of the existence of leaks before breaks must be obtained with a high degree of certainty, which requires large investments in a certain number of fields. The exact methodology will be defined by the Vendor and the utility at a later stage. However, from our point of view, analysis should involve the following steps: Definition of a maximum size defect; analysis of evolution under normal loads; stability analysis, definition of the critical defect; Definition of the detectable defect. This paper develops each of these points, and specifies, for each discipline, the domains in which present knowledge must be expanded and deepened [fr

Bayesian approach for the reliability assessment of corroded interdependent pipe networks

International Nuclear Information System (INIS)

Ait Mokhtar, El Hassene; Chateauneuf, Alaa; Laggoune, Radouane

2016-01-01

Pipelines under corrosion are subject to various environment conditions, and consequently it becomes difficult to build realistic corrosion models. In the present work, a Bayesian methodology is proposed to allow for updating the corrosion model parameters according to the evolution of environmental conditions. For reliability assessment of dependent structures, Bayesian networks are used to provide interesting qualitative and quantitative description of the information in the system. The qualitative contribution lies in the modeling of complex system, composed by dependent pipelines, as a Bayesian network. The quantitative one lies in the evaluation of the dependencies between pipelines by the use of a new method for the generation of conditional probability tables. The effectiveness of Bayesian updating is illustrated through an application where the new reliability of degraded (corroded) pipe networks is assessed. - Highlights: • A methodology for Bayesian network modeling of pipe networks is proposed. • Bayesian approach based on Metropolis - Hastings algorithm is conducted for corrosion model updating. • The reliability of corroded pipe network is assessed by considering the interdependencies between the pipelines.

Experimental study on fundamental phenomena in HTGR small break air-ingress accident

International Nuclear Information System (INIS)

Kim, Jae Soon; Hwang, Jin-Seok; Kim, Eung Soo; Kim, Byung Jun; Oh, Chang Ho

2016-01-01

Highlights: • Air-ingress phenomena on the small break in a HTGR are experimentally investigated. • Experiment is investigated for various break sizes, angles, and density ratios. • Maximum air-ingress rate is observed at 120° in break angle. • This study reveals that air-ingress in the small break is governed by; buoyancy and flow inertia. • A non-dimensional parameter is newly proposed to determine the air-ingress flow regimes. • Newly proposed parameter is based on buoyancy versus inertia force. - Abstract: This study experimentally investigates fundamental phenomena in the HTGR small break air-ingress accident. Several important parameters including density ratio, break angle, break size, and main flow velocity are considered in the measurement and the analysis. The test-section is made of a circular pipe with small holes drilled around the surface and it is installed in the helium/air flow circulation loop. Oxygen concentrations and flow rates are recorded during the tests with fixed break angles, break sizes, and flow velocities for measurement of the air-ingress rates. According to the experimental results, the higher density difference leads to the higher rates of air-ingress with large sensitivity of the break angles. It is also found that the break angle significantly affects the air-ingress rates, which is gradually increased from 0° to 120° and suddenly decreased to 180°. The minimum air ingress rate is found at 0° and the maximum, at 110°. The air-ingress rate increases with the break size due to the increased flow-exchange area. However, it is not directly proportional to the break area due to the complexity of the phenomena. The increased flow velocity in the channel inside enhances the air-ingress process. However, among all the parameters, the main flow velocity exhibits the lowest effect on this process. In this study, the Froude Number relevant to the small break air-ingress conditions are newly defined considering both heavy

Piping equipment; Materiel petrole

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This 'blue bible' of the perfect piping-man appeals to end-users of industrial facilities of the petroleum and chemical industries (purchase services, standardization, new works, maintenance) but also to pipe-makers and hollow-ware makers. It describes the characteristics of materials (carbon steels, stainless steels, alloyed steels, special alloys) and the dimensions of pipe elements: pipes, welding fittings, flanges, sealing products, forged steel fittings, forged steel valves, cast steel valves, ASTM standards, industrial valves. (J.S.)

Effects of off-centered crack and restraint of induced bending due to pressure on the crack-opening-area analysis of pipes

International Nuclear Information System (INIS)

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

1995-01-01

Estimation of leak rate is an important element in developing leak-before-break (LBB) methodology for piping integrity and safety analysis of nuclear power plants. Here, current models for the crack-opening-area analysis of pipes with circumferential through-wall cracks are based on various idealizations or assumption which are often necessary to simplify the mathematical formulation and numerical calculation. This paper focuses on the validity of two such assumptions involving off-centered cracks and restraint of induced bending due to pressure and quantifies their effects on the crack-opening analysis of pipes. Both finite element and/or simple estimation methods were employed to compute the center-crack-opening displacement and crack-opening shape for a through-wall-cracked pipe considering off-centered cracks and restrain of induced bending due to pressure. The results of analyses show that for both cases the crack-opening area can be reduced significantly. For pipes with off-centered cracks, the crack-opening area can be evaluated from analyses of symmetrically centered cracks and assuming elliptical profile. For pipes with complete restraint of induced bending due to pressure, the reduction of crack-opening area depends on the crack size. When the crack size is small, the restraint effects can be ignored. However, when the crack size is larger, the restrained crack-opening can be significantly smaller than the unrestrained crack-opening depending on the length of pipe involved, and hence, may be important for the crack-opening-area and leak-rate analyses

A Markov chain model for CANDU feeder pipe degradation

International Nuclear Information System (INIS)

Datla, S.; Dinnie, K.; Usmani, A.; Yuan, X.-X.

2008-01-01

There is need for risk based approach to manage feeder pipe degradation to ensure safe operation by minimizing the nuclear safety risk. The current lack of understanding of some fundamental degradation mechanisms will result in uncertainty in predicting the rupture frequency. There are still concerns caused by uncertainties in the inspection techniques and engineering evaluations which should be addressed in the current procedures. A probabilistic approach is therefore useful in quantifying the risk and also it provides a tool for risk based decision making. This paper discusses the application of Markov chain model for feeder pipes in order to predict and manage the risks associated with the existing and future aging-related feeder degradation mechanisms. The major challenge in the approach is the lack of service data in characterizing the transition probabilities of the Markov model. The paper also discusses various approaches in estimating plant specific degradation rates. (author)

Characteristics of iron corrosion scales and water quality variations in drinking water distribution systems of different pipe materials.

Science.gov (United States)

Li, Manjie; Liu, Zhaowei; Chen, Yongcan; Hai, Yang

2016-12-01

Interaction between old, corroded iron pipe surfaces and bulk water is crucial to the water quality protection in drinking water distribution systems (WDS). Iron released from corrosion products will deteriorate water quality and lead to red water. This study attempted to understand the effects of pipe materials on corrosion scale characteristics and water quality variations in WDS. A more than 20-year-old hybrid pipe section assembled of unlined cast iron pipe (UCIP) and galvanized iron pipe (GIP) was selected to investigate physico-chemical characteristics of corrosion scales and their effects on water quality variations. Scanning Electron Microscope (SEM), Energy Dispersive X-ray Spectroscopy (EDS), Inductively Coupled Plasma (ICP) and X-ray Diffraction (XRD) were used to analyze micromorphology and chemical composition of corrosion scales. In bench testing, water quality parameters, such as pH, dissolved oxygen (DO), oxidation reduction potential (ORP), alkalinity, conductivity, turbidity, color, Fe 2+ , Fe 3+ and Zn 2+ , were determined. Scale analysis and bench-scale testing results demonstrated a significant effect of pipe materials on scale characteristics and thereby water quality variations in WDS. Characteristics of corrosion scales sampled from different pipe segments show obvious differences, both in physical and chemical aspects. Corrosion scales were found highly amorphous. Thanks to the protection of zinc coatings, GIP system was identified as the best water quality stability, in spite of high zinc release potential. It is deduced that the complicated composition of corrosion scales and structural break by the weld result in the diminished water quality stability in HP system. Measurement results showed that iron is released mainly in ferric particulate form. Copyright © 2016 Elsevier Ltd. All rights reserved.

Heat Pipes

Science.gov (United States)

1990-01-01

Bobs Candies, Inc. produces some 24 million pounds of candy a year, much of it 'Christmas candy.' To meet Christmas demand, it must produce year-round. Thousands of cases of candy must be stored a good part of the year in two huge warehouses. The candy is very sensitive to temperature. The warehouses must be maintained at temperatures of 78-80 degrees Fahrenheit with relative humidities of 38- 42 percent. Such precise climate control of enormous buildings can be very expensive. In 1985, energy costs for the single warehouse ran to more than 57,000 for the year. NASA and the Florida Solar Energy Center (FSEC) were adapting heat pipe technology to control humidity in building environments. The heat pipes handle the jobs of precooling and reheating without using energy. The company contacted a FSEC systems engineer and from that contact eventually emerged a cooperative test project to install a heat pipe system at Bobs' warehouses, operate it for a period of time to determine accurately the cost benefits, and gather data applicable to development of future heat pipe systems. Installation was completed in mid-1987 and data collection is still in progress. In 1989, total energy cost for two warehouses, with the heat pipes complementing the air conditioning system was 28,706, and that figures out to a cost reduction.

Generation of plastic influence functions for J-integral and crack opening displacement of thin-walled pipes with a short circumferential through-wall crack

International Nuclear Information System (INIS)

Park, Jeong Soon; Choi, Young Hwan; Im, Seyoung

2014-01-01

Fracture mechanics parameters such as the J-integral and crack opening displacement (COD), are necessary for Leak-Before-Break (LBB) evaluation. The famous two estimation methods, the GE/EPRI and the Reference Stress Method (RSM), have their applicability limit with regard to the ratio of a pipe mean radius to thickness (R m /t). In order to extend their applicability limit to a thin walled pipe, several finite element analyses are performed for the J-integral and COD, and then new plastic influence functions are developed for thin-walled pipes with a short circumferential through-wall crack. With the newly generated plastic influence functions, the GE/EPRI and the RSM give closer results with those obtained from detailed finite element analyses. In addition, C*-integral and COD rate are estimated by using the new plastic influence functions and they are well matched with elastic–creep finite element analysis results under the power-law creep condition. Since the LBB concept can be applied to a piping system in a Korean Sodium-cooled Fast Reactor (SFR) which is designed to have thin-walled pipes and to operate in high temperature enough to cause creep, this paper can be applied for the LBB assessment of thin-walled pipes with a short through-wall crack in the SFR

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

User's manuals of probabilistic fracture mechanics analysis code for aged piping, PASCAL-SP

International Nuclear Information System (INIS)

Itoh, Hiroto; Nishikawa, Hiroyuki; Onizawa, Kunio; Kato, Daisuke; Osakabe, Kazuya

2010-03-01

As a part of research on the material degradation and structural integrity assessment for aged LWR components, a PFM (Probabilistic Fracture Mechanics) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed. This code evaluates the failure probabilities at welded joints of aged piping by a Monte Carlo method. PASCAL-SP treats stress corrosion cracking (SCC) and fatigue crack growth in piping, according to the approaches of NISA and JSME FFS Code. The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the latest knowledge in the SCC assessment and fracture criteria of piping. In addition, the accuracy of flaw detection and sizing at in-service inspection and residual stress distribution were modeled based on experimental data and introduced into PASCAL-SP. This code has been developed for a cross-check use by the regulatory body in Japan. In addition to this, this code can also be used for a research purpose by researchers in academia and industries. This report provides the user's manual and theoretical background of the code. (author)

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

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

Microstructural characterization of primary coolant pipe steel

International Nuclear Information System (INIS)

Miller, M.K.; Bentley, J.

1986-01-01

Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

Pipe rupture test results; 4 inch pipe whip tests under BWR operational condition-clearance parameter experiments

International Nuclear Information System (INIS)

Ueda, Syuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kurihara, Ryoichi; Kato, Rokuro; Saito, Kazuo; Miyazono, Shohachiro

1981-05-01

The purpose of pipe rupture studies in JAERI is to perform the model tests on pipe whip, restraint behavior, jet impingement and jet thrust force, and to establish the computational method for analyzing these phenomena. This report describes the experimental results of pipe whip on the pipe specimens of 4 inch in diameter under BWR condition on which the pressure is 6.77 MPa and the temperature is 285 0 C. The pipe specimens were 114.3 mm (4 inch) in diameter and 8.6 mm in thickness and 4500 mm in length. Four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from type 304 stainless steel. The experimental parameter was the clearance (30, 50 and 100 mm). The dynamic strain behavior of the pipe specimen and the restraints was investigated by strain gages and their residual deformation was obtained by measuring marking points provided on their surface. The Pressure-time history in the pipe specimens was also obtained by pressure gages. The maximum pipe strain is caused near the restraints and increases with increase of the clearance. The experimental results of pipe whip tests indicate the effectiveness of pipe whip restraints. The ratio of absorbed strain energy of the pipe specimen to that of the restraints is nearly constant for different clearances at the overhang length of 400 mm. (author)

A new mechanism for periodic bursting of the recirculation region in the flow through a sudden expansion in a circular pipe

Science.gov (United States)

Lebon, Benoit; Nguyen, Minh Quan; Peixinho, Jorge; Shadloo, Mostafa Safdari; Hadjadj, Abdellah

2018-03-01

We report the results of a combined experimental and numerical study of specific finite-amplitude disturbances for transition to turbulence in the flow through a circular pipe with a sudden expansion. The critical amplitude thresholds for localized turbulent patch downstream of the expansion scale with the Reynolds number with a power law exponent of -2.3 for experiments and -2.8 for simulations. A new mechanism for the periodic bursting of the recirculation region is uncovered where the asymmetric recirculation flow develops a periodic dynamics: a secondary recirculation breaks the symmetry along the pipe wall and bursts into localized turbulence, which travels downstream and relaminarises. Flow visualizations show a simple flow pattern of three waves forming, growing, and bursting.

A non extensive approach for DNA breaking by ionizing radiation

OpenAIRE

Sotolongo-Costa, O; Guzman, F; Antoranz, JC; Rodgers, GJ; Rodriguez, O; Arruda Neto, JDT; Deepman, A

2002-01-01

Tsallis entropy and a maximum entropy principle allows to reproduce experimental data of DNA double strand breaking by electron and neutron radiation. Analytic results for the probability of finding a DNA segment of length l are obtained reproducing quite well the fragment distribution function experimentally obtained.

Saturation behaviour of the LHC NEG coated beam pipes

CERN Document Server

Porcelli, T; Lanza, G; Baglin, V; Jimenez, J M

2012-01-01

In the CERN Large Hadron Collider (LHC), about 6 km of the UHV beam pipe are at room temperature and serve as experimental or utility insertions. TiZrV non-evaporable getter (NEG) coating is used to maintain the design pressure during beam operation. Molecular desorption due to dynamic effects is stimulated during protons operation at high intensity. This phenomenon produces an important gas load from the vacuum chamber walls, which could lead to a partial or total saturation of the NEG coating. To keep the design vacuum performances and to schedule technical interventions for NEG reactivation, it is necessary to take into account all these aspects and to regularly evaluate the saturation level of the NEG coating. Experimental studies of a typical LHC vacuum sector were conducted in the laboratory in order to identify the best method to assess the saturation level of the beam pipe. Partial saturation of the NEG was performed and the effective pumping speed, transmission and capture probability are analysed.

Consistency of Trend Break Point Estimator with Underspecified Break Number

Directory of Open Access Journals (Sweden)

Jingjing Yang

2017-01-01

Full Text Available This paper discusses the consistency of trend break point estimators when the number of breaks is underspecified. The consistency of break point estimators in a simple location model with level shifts has been well documented by researchers under various settings, including extensions such as allowing a time trend in the model. Despite the consistency of break point estimators of level shifts, there are few papers on the consistency of trend shift break point estimators in the presence of an underspecified break number. The simulation study and asymptotic analysis in this paper show that the trend shift break point estimator does not converge to the true break points when the break number is underspecified. In the case of two trend shifts, the inconsistency problem worsens if the magnitudes of the breaks are similar and the breaks are either both positive or both negative. The limiting distribution for the trend break point estimator is developed and closely approximates the finite sample performance.

Heat pipe applications workshop report

International Nuclear Information System (INIS)

Ranken, W.A.

1978-04-01

The proceedings of the Heat Pipe Applications Workshop, held at the Los Alamos Scientific Laboratory October 20-21, 1977, are reported. This workshop, which brought together representatives of the Department of Energy and of a dozen industrial organizations actively engaged in the development and marketing of heat pipe equipment, was convened for the purpose of defining ways of accelerating the development and application of heat pipe technology. Recommendations from the three study groups formed by the participants are presented. These deal with such subjects as: (1) the problem encountered in obtaining support for the development of broadly applicable technologies, (2) the need for applications studies, (3) the establishment of a heat pipe technology center of excellence, (4) the role the Department of Energy might take with regard to heat pipe development and application, and (5) coordination of heat pipe industry efforts to raise the general level of understanding and acceptance of heat pipe solutions to heat control and transfer problems

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)

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)

Introduction to Heat Pipes

Science.gov (United States)

Ku, Jentung

2015-01-01

This is the presentation file for the short course Introduction to Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. NCTS 21070-15. Course Description: This course will present operating principles of the heat pipe with emphases on the underlying physical processes and requirements of pressure and energy balance. Performance characterizations and design considerations of the heat pipe will be highlighted. Guidelines for thermal engineers in the selection of heat pipes as part of the spacecraft thermal control system, testing methodology, and analytical modeling will also be discussed.

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)

Effects of Schmidt number on near-wall turbulent mass transfer in pipe flow

Energy Technology Data Exchange (ETDEWEB)

Kang, Chang Woo; Yang, Kyung Soo [Inha University, Incheon (Korea, Republic of)

2014-12-15

Large Eddy simulation (LES) of turbulent mass transfer in circular-pipe flow has been performed to investigate the characteristics of turbulent mass transfer in the near-wall region. We consider a fully-developed turbulent pipe flow with a constant wall concentration. The Reynolds number under consideration is Re{sub r} = 500 based on the friction velocity and the pipe radius, and the selected Schmidt numbers (Sc) are 0.71, 5, 10, 20 and 100. Dynamic subgrid-scale (SGS) models for the turbulent SGS stresses and turbulent mass fluxes were employed to close the governing equations. The current paper reports a comprehensive characterization of turbulent mass transfer in circular-pipe flow, focusing on its near-wall characteristics and Sc dependency. We start with mean fields by presenting mean velocity and concentration profiles, mean Sherwood numbers and mean mass transfer coefficients for the selected values of the parameters. After that, we present the characteristics of fluctuations including root-mean-square (rms) profiles of velocity, concentration, and mass transfer coefficient fluctuations. Turbulent mass fluxes and correlations between velocity and concentration fluctuations are also discussed. The near-wall behaviour of turbulent diffusivity and turbulent Schmidt number is shown, and other authors' correlations on their limiting behaviour towards the pipe wall are evaluated based on our LES results. The intermittent characteristics of turbulent mass transfer in pipe flow are depicted by probability density functions (pdf) of velocity and concentration fluctuations; joint pdfs between them are also presented. Instantaneous snapshots of velocity and concentration fluctuations are shown to supplement our discussion on the turbulence statistics. Finally, we report the results of octant analysis and budget calculation of concentration variance to clarify Sc-dependency of the correlation between near-wall turbulence structures and concentration fluctuation in

Effects of Schmidt number on near-wall turbulent mass transfer in pipe flow

International Nuclear Information System (INIS)

Kang, Chang Woo; Yang, Kyung Soo

2014-01-01

Large Eddy simulation (LES) of turbulent mass transfer in circular-pipe flow has been performed to investigate the characteristics of turbulent mass transfer in the near-wall region. We consider a fully-developed turbulent pipe flow with a constant wall concentration. The Reynolds number under consideration is Re r = 500 based on the friction velocity and the pipe radius, and the selected Schmidt numbers (Sc) are 0.71, 5, 10, 20 and 100. Dynamic subgrid-scale (SGS) models for the turbulent SGS stresses and turbulent mass fluxes were employed to close the governing equations. The current paper reports a comprehensive characterization of turbulent mass transfer in circular-pipe flow, focusing on its near-wall characteristics and Sc dependency. We start with mean fields by presenting mean velocity and concentration profiles, mean Sherwood numbers and mean mass transfer coefficients for the selected values of the parameters. After that, we present the characteristics of fluctuations including root-mean-square (rms) profiles of velocity, concentration, and mass transfer coefficient fluctuations. Turbulent mass fluxes and correlations between velocity and concentration fluctuations are also discussed. The near-wall behaviour of turbulent diffusivity and turbulent Schmidt number is shown, and other authors' correlations on their limiting behaviour towards the pipe wall are evaluated based on our LES results. The intermittent characteristics of turbulent mass transfer in pipe flow are depicted by probability density functions (pdf) of velocity and concentration fluctuations; joint pdfs between them are also presented. Instantaneous snapshots of velocity and concentration fluctuations are shown to supplement our discussion on the turbulence statistics. Finally, we report the results of octant analysis and budget calculation of concentration variance to clarify Sc-dependency of the correlation between near-wall turbulence structures and concentration fluctuation in the

Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings

International Nuclear Information System (INIS)

Brekow, G.; Wuestenberg, H.; Hesselmann, H.; Rathgeb, W.

1991-01-01

Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings. The pipe connection inner corner tests in feedwater lines to the main coolant pipe were carried out by Preussen-Elektra in cooperation with Siemens KWU and the BAM with the ultrasonic phased array method. The testing plan was developed by means of a computed model. For a trial of the testing plan, numerous ultrasonic measurements with the phased array method were carried out using a pipe test piece with TH-type inner edges, which was a 1:1 model of the reactor component to be tested. The data measured at several test notches in the pipe connection inner edge area covered by a plating of 6 mm were analyzed. (orig./MM) [de

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

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

Unification of SUSY breaking and GUT breaking

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tatsuo [Department of Physics, Hokkaido University,Sapporo 060-0810 (Japan); Omura, Yuji [Department of Physics, Nagoya University,Nagoya 464-8602 (Japan)

2015-02-18

We build explicit supersymmetric unification models where grand unified gauge symmetry breaking and supersymmetry (SUSY) breaking are caused by the same sector. Besides, the SM-charged particles are also predicted by the symmetry breaking sector, and they give the soft SUSY breaking terms through the so-called gauge mediation. We investigate the mass spectrums in an explicit model with SU(5) and additional gauge groups, and discuss its phenomenological aspects. Especially, nonzero A-term and B-term are generated at one-loop level according to the mediation via the vector superfields, so that the electro-weak symmetry breaking and 125 GeV Higgs mass may be achieved by the large B-term and A-term even if the stop mass is around 1 TeV.

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.)

Fabrication of a multi-walled metal pipe

International Nuclear Information System (INIS)

Shimamune, Koji; Toda, Saburo; Ishida, Ryuichi; Hatanaka, Tatsuo.

1969-01-01

In concentrically arranged metal pipes for simulated fuel elements in the form of a multi-walled pipe, their one end lengthens gradually in the axial direction from inner and outer pipes toward a central pipe for easy adjustment of deformation which occurs when the pipes are drawn. A plastic electrical insulator is disposed between adjacent pipes. Each end of the pipes is equipped with an annular flexible stopper which is allowed to travel in the axial direction so as to prevent the insulator from falling during drawing work. At the other end, all pipes are constricted and joined to each other to thereby form the desired multi-walled pipe. (Mikami, T.)

Single-strand breaks induced in Bacillus subtilis DNA by ultraviolet light: action spectrum and properties

International Nuclear Information System (INIS)

Peak, M.J.; Peak, J.G.

1982-01-01

The induction of single-strand breaks (alkali-labile bonds plus frank breaks) in the DNA of Bacillus subtilis irradiated in vivo by monochromatic UV light at wavelengths from 254 to 434nm was measured. The spectrum consists of a major far-UV (below 320nm) component and a minor near-UV shoulder. A mutant deficient in DNA polymerase I accumulates breaks caused by near-UV (above 320nm) wavelengths faster than the wild-type strain proficient in polymerase I. Measurable breaks in extracted DNA are induced at a higher frequency than those induced in vivo. Anoxia, glycerol, and diazobicyclo (2.2.2.) octane inhibit break formation in extracted DNA. Alkali-labile bonds induced by 365-nm UV radiation are largely (78%) covalent bond chain breaks, the remainder consists of true alkali-labile bonds, probably apurinic and apyrimidinic sites. (author)

Belgian experience in applying the open-quotes leak-before-breakclose quotes concept to the primary loop piping

International Nuclear Information System (INIS)

Gerard, R.; Malekian, C.; Meessen, O.

1997-01-01

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the open-quote two cutsclose quotes technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented

Reliability analysis of stiff versus flexible piping

International Nuclear Information System (INIS)

Lu, S.C.

1985-01-01

The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. The authors then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. They concluded that these changes substantially reduce calculated piping responses and allow piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design. An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Results indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design

Corroded scale analysis from water distribution pipes

Directory of Open Access Journals (Sweden)

Rajaković-Ognjanović Vladana N.

2011-01-01

Full Text Available The subject of this study was the steel pipes that are part of Belgrade's drinking water supply network. In order to investigate the mutual effects of corrosion and water quality, the corrosion scales on the pipes were analyzed. The idea was to improve control of corrosion processes and prevent impact of corrosion on water quality degradation. The instrumental methods for corrosion scales characterization used were: scanning electron microscopy (SEM, for the investigation of corrosion scales of the analyzed samples surfaces, X-ray diffraction (XRD, for the analysis of the presence of solid forms inside scales, scanning electron microscopy (SEM, for the microstructural analysis of the corroded scales, and BET adsorption isotherm for the surface area determination. Depending on the composition of water next to the pipe surface, corrosion of iron results in the formation of different compounds and solid phases. The composition and structure of the iron scales in the drinking water distribution pipes depends on the type of the metal and the composition of the aqueous phase. Their formation is probably governed by several factors that include water quality parameters such as pH, alkalinity, buffer intensity, natural organic matter (NOM concentration, and dissolved oxygen (DO concentration. Factors such as water flow patterns, seasonal fluctuations in temperature, and microbiological activity as well as water treatment practices such as application of corrosion inhibitors can also influence corrosion scale formation and growth. Therefore, the corrosion scales found in iron and steel pipes are expected to have unique features for each site. Compounds that are found in iron corrosion scales often include goethite, lepidocrocite, magnetite, hematite, ferrous oxide, siderite, ferrous hydroxide, ferric hydroxide, ferrihydrite, calcium carbonate and green rusts. Iron scales have characteristic features that include: corroded floor, porous core that contains

More dynamical supersymmetry breaking

International Nuclear Information System (INIS)

Csaki, C.; Randall, L.; Skiba, W.

1996-01-01

In this paper we introduce a new class of theories which dynamically break supersymmetry based on the gauge group SU(n) x SU(3) x U(1) for even n. These theories are interesting in that no dynamical superpotential is generated in the absence of perturbations. For the example SU(4) x SU(3) x U(1) we explicitly demonstrate that all flat directions can be lifted through a renormalizable superpotential and that supersymmetry is dynamically broken. We derive the exact superpotential for this theory, which exhibits new and interesting dynamical phenomena. For example, modifications to classical constraints can be field dependent. We also consider the generalization to SU(n) x SU(3) x U(1) models (with even n>4). We present a renormalizable superpotential which lifts all flat directions. Because SU(3) is not confining in the absence of perturbations, the analysis of supersymmetry breaking is very different in these theories from the n=4 example. When the SU(n) gauge group confines, the Yukawa couplings drive the SU(3) theory into a regime with a dynamically generated superpotential. By considering a simplified version of these theories we argue that supersymmetry is probably broken. (orig.)

Heat pipes in modern heat exchangers

International Nuclear Information System (INIS)

Vasiliev, Leonard L.

2005-01-01

Heat pipes are very flexible systems with regard to effective thermal control. They can easily be implemented as heat exchangers inside sorption and vapour-compression heat pumps, refrigerators and other types of heat transfer devices. Their heat transfer coefficient in the evaporator and condenser zones is 10 3 -10 5 W/m 2 K, heat pipe thermal resistance is 0.01-0.03 K/W, therefore leading to smaller area and mass of heat exchangers. Miniature and micro heat pipes are welcomed for electronic components cooling and space two-phase thermal control systems. Loop heat pipes, pulsating heat pipes and sorption heat pipes are the novelty for modern heat exchangers. Heat pipe air preheaters are used in thermal power plants to preheat the secondary-primary air required for combustion of fuel in the boiler using the energy available in exhaust gases. Heat pipe solar collectors are promising for domestic use. This paper reviews mainly heat pipe developments in the Former Soviet Union Countries. Some new results obtained in USA and Europe are also included

Application of the leak-before-break concept to steam generator tubes

International Nuclear Information System (INIS)

Keim, E.; Kastner, W.

1994-01-01

The Leak-Before-Break (LBB) behaviour of a piping component means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safety 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. The fracture mechanics analysis supplies the input for the thermal-hydraulic analysis. The resulting leakage rate related to the crack length of a longitudinal or circumferential crack and the minimum detectable values of leakage rate and crack length lead to two criteria, which allow for the LBB-behaviour of the pipe: - the critical crack length must be larger than the crack length being safety detected by leakage monitoring systems (LMS) - the critical crack length must be larger than the crack length being safety detected by non-destructive examination (NDE). This LBB-concept is applied to steam generator (SG) tubes. Two examples, which will be presented, show that this concept is a very useful and effective tool which allows the prediction of LBB-behaviour of SG tubes. (Author)

Leakage Detection and Estimation Algorithm for Loss Reduction in Water Piping Networks

Directory of Open Access Journals (Sweden)

Kazeem B. Adedeji

2017-10-01

Full Text Available Water loss through leaking pipes constitutes a major challenge to the operational service of water utilities. In recent years, increasing concern about the financial loss and environmental pollution caused by leaking pipes has been driving the development of efficient algorithms for detecting leakage in water piping networks. Water distribution networks (WDNs are disperse in nature with numerous number of nodes and branches. Consequently, identifying the segment(s of the network and the exact leaking pipelines connected to this segment(s where higher background leakage outflow occurs is a challenging task. Background leakage concerns the outflow from small cracks or deteriorated joints. In addition, because they are diffuse flow, they are not characterised by quick pressure drop and are not detectable by measuring instruments. Consequently, they go unreported for a long period of time posing a threat to water loss volume. Most of the existing research focuses on the detection and localisation of burst type leakages which are characterised by a sudden pressure drop. In this work, an algorithm for detecting and estimating background leakage in water distribution networks is presented. The algorithm integrates a leakage model into a classical WDN hydraulic model for solving the network leakage flows. The applicability of the developed algorithm is demonstrated on two different water networks. The results of the tested networks are discussed and the solutions obtained show the benefits of the proposed algorithm. A noteworthy evidence is that the algorithm permits the detection of critical segments or pipes of the network experiencing higher leakage outflow and indicates the probable pipes of the network where pressure control can be performed. However, the possible position of pressure control elements along such critical pipes will be addressed in future work.

Flow in a circular expansion pipe flow: effect of a vortex perturbation on localised turbulence

Energy Technology Data Exchange (ETDEWEB)

Selvam, Kamal; Peixinho, Jorge [Laboratoire Ondes Milieux Complexes, CNRS and Université du Havre, F-76600 Le Havre (France); Willis, Ashley P, E-mail: jorge.peixinho@univ-lehavre.fr [School of Mathematics and Statistics, University of Sheffield, Sheffield S3 7RH (United Kingdom)

2016-12-15

We report the results of three-dimensional direct numerical simulations for incompressible viscous fluid in a circular pipe flow with a sudden expansion. At the inlet, a parabolic velocity profile is applied together with a finite amplitude perturbation in the form of a vortex with its axis parallel to the axis of the pipe. At sufficiently high Reynolds numbers the recirculation region breaks into a turbulent patch that changes position axially, depending on the strength of the perturbation. This vortex perturbation is believed to produce a less abrupt transition than in previous studies, which applied a tilt perturbation, as the localised turbulence is observed via the formation of a wavy structure at a low order azimuthal mode, which resembles an optimally amplified perturbation. For large vortex amplitude, the localised turbulence remains at a constant axial position. It is further investigated using proper orthogonal decomposition, which indicates that the centre region close to the expansion is highly energetic. (paper)

Modeling impurity-assisted chain creation in noble-metal break junctions

International Nuclear Information System (INIS)

Di Napoli, S; Thiess, A; Blügel, S; Mokrousov, Y

2012-01-01

In this work we present the generalization of the model for chain formation in break junctions, introduced by Thiess et al (2008 Nano Lett. 8 2144), to zigzag transition-metal chains with s and p impurities. We apply this extended model to study the producibility trends for noble-metal chains with impurities, often present in break junction experiments, namely, Cu, Ag and Au chains with H, C, O and N adatoms. Providing the material-specific parameters for our model from systematic full-potential linearized augmented plane-wave first-principles calculations, we find that the presence of such impurities crucially affects the binding properties of the noble-metal chains. We reveal that both the impurity-induced bond strengthening and the formation of zigzag bonds can lead to a significantly enhanced probability for chain formation in break junctions. (paper)

46 CFR 61.15-5 - Steam piping.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Steam piping. 61.15-5 Section 61.15-5 Shipping COAST... Periodic Tests of Piping Systems § 61.15-5 Steam piping. (a) Main steam piping shall be subjected to a... removed and the piping thoroughly examined. (b) All steam piping subject to pressure from the main boiler...

Factors controlling the spatial distribution of soil piping erosion on loess-derived soils: A case study from central Belgium

Science.gov (United States)

Verachtert, E.; Van Den Eeckhaut, M.; Poesen, J.; Deckers, J.

2010-06-01

Collapsible loess-derived soils are prone to soil piping erosion, where enlargement of macropores may lead to a subsurface pipe network and eventually to soil collapse and gully development. This study aims at understanding the main factors controlling spatial patterns of piping in loess-derived soils under a temperate climate. To map the spatial distribution of piping and identify the environmental controls on its distribution, a regional survey was carried out in a 236 km 2 study area in the Flemish Ardennes (Belgium). Orthophotos taken at optimal field conditions (winter) were analyzed to detect piping in open landscapes and ground thruthing was systematically done through field surveys. In total, 137 parcels having 560 collapsed pipes were mapped. Dimensions of the sinkholes and local slope gradient were measured in the field and topographical variables were derived from LiDAR data. Land use plays an important role as 97% of the sites with piping are found under pasture. The probability of piping increases rapidly on hillslopes with gradients exceeding 8% and with a concave profile and plan curvature, enhancing subsurface flow concentration. The zones with soil profiles on shallow loess over a relatively thin layer of homogeneous blue massive clays (Aalbeke Member) are most prone to piping. Soil characteristics are of less importance to explain piping occurrence. Furthermore, the topographical threshold line indicating the critical slope gradient for a given contributing drainage area was determined. This threshold line (negative power relation) is similar to the threshold line for shallow gully initiation.

Heat pipe development

Science.gov (United States)

Bienart, W. B.

1973-01-01

The objective of this program was to investigate analytically and experimentally the performance of heat pipes with composite wicks--specifically, those having pedestal arteries and screwthread circumferential grooves. An analytical model was developed to describe the effects of screwthreads and screen secondary wicks on the transport capability of the artery. The model describes the hydrodynamics of the circumferential flow in triangular grooves with azimuthally varying capillary menisci and liquid cross-sections. Normalized results were obtained which give the influence of evaporator heat flux on the axial heat transport capability of the arterial wick. In order to evaluate the priming behavior of composite wicks under actual load conditions, an 'inverted' glass heat pipe was designed and constructed. The results obtained from the analysis and from the tests with the glass heat pipe were applied to the OAO-C Level 5 heat pipe, and an improved correlation between predicted and measured evaporator and transport performance were obtained.

Linker-dependent Junction Formation Probability in Single-Molecule Junctions

Energy Technology Data Exchange (ETDEWEB)

Yoo, Pil Sun; Kim, Taekyeong [HankukUniversity of Foreign Studies, Yongin (Korea, Republic of)

2015-01-15

We compare the junction formation probabilities of single-molecule junctions with different linker molecules by using a scanning tunneling microscope-based break-junction technique. We found that the junction formation probability varies as SH > SMe > NH2 for the benzene backbone molecule with different types of anchoring groups, through quantitative statistical analysis. These results are attributed to different bonding forces according to the linker groups formed with Au atoms in the electrodes, which is consistent with previous works. Our work allows a better understanding of the contact chemistry in the metal.molecule junction for future molecular electronic devices.

A study on the effect of flaw detection probability assumptions on risk reduction achieved by non-destructive inspection

International Nuclear Information System (INIS)

Cronvall, O.; Simola, K.; Männistö, I.; Gunnars, J.; Alverlind, L.; Dillström, P.; Gandossi, L.

2012-01-01

Leakages and ruptures of piping components lead to reduction or loss of the pressure retaining capability of the system, and thus contribute to the overall risk associated with nuclear power plants. In-service inspection (ISI) aims at verifying that defects are not present in components of the pressure boundary or, if defects are present, ensuring that these are detected before they affect the safe operation of the plant. Reliability estimates of piping are needed e.g., in probabilistic safety assessment (PSA) studies, risk-informed ISI (RI-ISI) applications, and other structural reliability assessments. Probabilistic fracture mechanics models can account for ISI reliability, but a quantitative estimate for the latter is needed. This is normally expressed in terms of probability of detection (POD) curves, which correlate the probability of detecting a flaw with flaw size. A detailed POD curve is often difficult (or practically impossible) to obtain. If sufficient risk reduction can be shown by using simplified (but reasonably conservative) POD estimates, more complex PODs are not needed. This paper summarises the results of a study on the effect of piping inspection reliability assumptions on failure probability using structural reliability models. The main interest was to investigate whether it is justifiable to use a simplified POD curve. Further, the study compared various structural reliability calculation approaches for a set of analysis cases. The results indicate that the use of a simplified POD could be justifiable in RI-ISI applications.

Replaceable liquid nitrogen piping

International Nuclear Information System (INIS)

Yasujima, Yasuo; Sato, Kiyoshi; Sato, Masataka; Hongo, Toshio

1982-01-01

This liquid nitrogen piping with total length of about 50 m was made and installed to supply the liquid nitrogen for heat insulating shield to three superconducting magnets for deflection and large super-conducting magnet for detection in the π-meson beam line used for high energy physics experiment in the National Laboratory for High Energy Physics. The points considered in the design and manufacture stages are reported. In order to minimize the consumption of liquid nitrogen during transport, vacuum heat insulation method was adopted. The construction period and cost were reduced by the standardization of the components, the improvement of welding works and the elimination of ineffective works. For simplifying the maintenance, spare parts are always prepared. The construction and the procedure of assembling of the liquid nitrogen piping are described. The piping is of double-walled construction, and its low temperature part was made of SUS 316L. The super-insulation by aluminum vacuum evaporation and active carbon were attached on the external surface of the internal pipe. The final leak test and the heating degassing were performed. The tests on evacuation, transport capacity and heat entry are reported. By making the internal pipe into smaller size, the piping may be more efficient. (Kako, I.)

Transients in pipes

International Nuclear Information System (INIS)

Marchesin, D.; Paes-Leme, P.J.S.; Sampaio, R.

1981-01-01

The motion of a fluid in a pipe is commonly modeled utilizing the one space dimension conservation laws of mass and momentum. The development of shocks and spikes utilizing the uniform sampling method is studied. The effects of temperature variations and friction are compared for gas pipes. (Author) [pt

Current results for the NRC's short cracks in piping and piping welds research program

International Nuclear Information System (INIS)

Wilkowski, G.; Krishnaswamy, P. Brust, F.; Francini, R.; Ghadiali, N.; Kilinski, T.; Marschall, C.; Rahman, S.; Rosenfield, A.; Scott, P.

1994-01-01

The overall objective of the Short Cracks in Piping and Piping Welds Program is to verify and improve engineering analyses to predict the fracture behavior of circumferentially cracked pipe under quasi-static loading with particular attention to crack lengths typically used in LBB or flaw evaluation criteria. The program consists of 8 technical tasks as listed below. Task 1 Short through-wall-cracked (TWC) pipe evaluations. Task 2 Short surface-cracked pipe evaluations. Task 3 Bi-metallic weld crack evaluations. Task 4 Dynamic strain aging and crack instabilities. Task 5 Fracture evaluations of anisotropic pipe. Task 6 Crack-opening-area evaluations. Task 7 NRCPIPE Code improvements. Task 8 Additional efforts. Since the last WRSM meeting several additional tasks have been initiated in this program. These are discussed in Task 8. Based on results to date, the first seven tasks have also been modified as deemed necessary. The most significant accomplishments in each of these tasks since the last WRSIM meeting are discussed below. The details of all the results presented here are published in the semiannual reports from this program

PPOOLEX experiments with two parallel blowdown pipes

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

2011-01-15

This report summarizes the results of the experiments with two transparent blowdown pipes carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through either one or two vertical transparent blowdown pipes to the condensation pool. Five experiments with one pipe and six with two parallel pipes were carried out. The main purpose of the experiments was to study loads caused by chugging (rapid condensation) while steam is discharged into the condensation pool filled with sub-cooled water. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. In the experiments the initial temperature of the condensation pool water varied from 12 deg. C to 55 deg. C, the steam flow rate from 40 g/s to 1 300 g/s and the temperature of incoming steam from 120 deg. C to 185 deg. C. In the experiments with only one transparent blowdown pipe chugging phenomenon didn't occur as intensified as in the preceding experiments carried out with a DN200 stainless steel pipe. With the steel blowdown pipe even 10 times higher pressure pulses were registered inside the pipe. Meanwhile, loads registered in the pool didn't indicate significant differences between the steel and polycarbonate pipe experiments. In the experiments with two transparent blowdown pipes, the steamwater interface moved almost synchronously up and down inside both pipes. Chugging was stronger than in the one pipe experiments and even two times higher loads were measured inside the pipes. The loads at the blowdown pipe outlet were approximately the same as in the one pipe cases. Other registered loads around the pool were about 50-100 % higher than with one pipe. The experiments with two parallel blowdown pipes gave contradictory results compared to the earlier studies dealing with chugging loads in case of multiple pipes. Contributing

Automated ultrasonic pipe weld inspection. Part 1

International Nuclear Information System (INIS)

Karl Deutsch, W.A.; Schulte, P.; Joswig, M.; Kattwinkel, R.

2006-01-01

This article contains a brief overview on automated ultrasonic welded inspection for various pipe types. Some inspection steps might by carried out with portable test equipment (e.g. pipe and test), but the weld inspection in all internationally relevant specification must be automated. The pipe geometry, the production process, and the pipe usage determine the number of required probes. Recent updates for some test specifications enforce a large number of ultrasonic probes, e.g. the Shell standard. Since seamless pipes are sometimes replaced by ERW pipes and LSAW pipes (in both cases to save production cost), the inspection methods change gradually between the various pipe types. Each testing system is unique and shows its specialties which have to be discussed by supplier, testing system user and final customer of the pipe. (author)

Loss-of-coolant accident for large pipe breaks in light water reactor plants

International Nuclear Information System (INIS)

Keusenhoff, J.

1980-01-01

The importance of loss-of-coolant accidents (LOCA) and their control for nuclear reactor safety is explained. Showing the cooling circuits and emergency core cooling systems (ECCS) of both, PWR and BWR, the possible break spectrum and the general sequence of events is discussed. The governing physical phenomena for the different LOCA phases are pointed out in more detail. Special emphasis is taken on rules, regulations and failure criteria for licensing purposes. Analysis methods and codes for both, evaluation and best-estimate model are compared under deterministic and probabilistic approach, respectively. Some insight in present integral and separate effect tests demonstrates the interdependency of analysis and experiment. Results of LOCA analysis and experiments show the present state of the art. (orig.)

Passive advection-dispersion in networks of pipes: Effect of connectivity and relationship to permeability

Science.gov (United States)

Bernabé, Y.; Wang, Y.; Qi, T.; Li, M.

2016-02-01

The main purpose of this work is to investigate the relationship between passive advection-dispersion and permeability in porous materials presumed to be statistically homogeneous at scales larger than the pore scale but smaller than the reservoir scale. We simulated fluid flow through pipe network realizations with different pipe radius distributions and different levels of connectivity. The flow simulations used periodic boundary conditions, allowing monitoring of the advective motion of solute particles in a large periodic array of identical network realizations. In order to simulate dispersion, we assumed that the solute particles obeyed Taylor dispersion in individual pipes. When a particle entered a pipe, a residence time consistent with local Taylor dispersion was randomly assigned to it. When exiting the pipe, the particle randomly proceeded into one of the pipes connected to the original one according to probabilities proportional to the outgoing volumetric flow in each pipe. For each simulation we tracked the motion of at least 6000 solute particles. The mean fluid velocity was 10-3 ms-1, and the distance traveled was on the order of 10 m. Macroscopic dispersion was quantified using the method of moments. Despite differences arising from using different types of lattices (simple cubic, body-centered cubic, and face-centered cubic), a number of general observations were made. Longitudinal dispersion was at least 1 order of magnitude greater than transverse dispersion, and both strongly increased with decreasing pore connectivity and/or pore size variability. In conditions of variable hydraulic radius and fixed pore connectivity and pore size variability, the simulated dispersivities increased as power laws of the hydraulic radius and, consequently, of permeability, in agreement with previously published experimental results. Based on these observations, we were able to resolve some of the complexity of the relationship between dispersivity and permeability.

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

Nitrogen heat pipe for cryocooler thermal shunt

International Nuclear Information System (INIS)

Prenger F.C.; Hill, D.D.; Daney, D.E.

1996-01-01

A nitrogen heat pipe was designed, built and tested for the purpose of providing a thermal shunt between the two stages of a Gifford-McMahan (GM) cryocooler during cooldown. The nitrogen heat pipe has an operating temperature range between 63 and 123 K. While the heat pipe is in this temperature range during the system cooldown, it acts as a thermal shunt between the first and second stage of the cryocooler. The heat pipe increases the heat transfer to the first stage of the cryocooler, thereby reducing the cooldown time of the system. When the heat pipe temperature drops below the triple point, the nitrogen working fluid freezes, effectively stopping the heat pipe operation. A small heat leak between cryocooler stages remains because of axial conduction along the heat pipe wall. As long as the heat pipe remains below 63 K, the heat pipe remains inactive. Heat pipe performance limits were measured and the optimum fluid charge was determined

Pipe grabber

Energy Technology Data Exchange (ETDEWEB)

Sharafutdinov, I.G.; Mubashirov, S.G.; Prokopov, O.I.

1981-05-15

A pipe grabber is suggested which contains a housing, clamping elements and centering mechanism with drive installed on the lower end of the housing. In order to improve the reliable operation of the pipe grabber, the centering mechanism is made in the form of a reinforced ringed flexible shaft, while the drive is made in the form of elastic rotating discs. In this case the direction of rotation of the discs and the flexible shaft is the opposite.

Advanced industrial ceramic heat pipe recuperators

Energy Technology Data Exchange (ETDEWEB)

Strumpf, H.J.; Stillwagon, T.L.; Kotchick, D.M.; Coombs, M.G.

1988-01-01

This paper summarizes the results of an investigation involving the use of ceramic heat pipe recuperators for high-temperature heat recovery from industrial furnaces. The function of the recuperator is to preheat combustion air with furnace exhaust gas. The heat pipe recuperator comprises a bundle of individual ceramic heat pipes acting in concert, with a partition separating the air and exhaust gas flow streams. Because each heat pipe is essentially an independent heat exchanger, the failure of a single tube does not compromise recuperator integrity, has only a minimal effect on overall heat exchanger performance and enables easier replacement of individual heat pipes. In addition, the heat pipe acts as an essentially isothermal heat transfer device, leading to a high thermodynamic efficiency. Cost estimates developed for heat pipe recuperator systems indicate favorable payback periods. Laboratory studies have demonstrated the feasibility of fabricating the required ceramic tubes, coating the inside of the tubes with CVD tungsten, and sealing the heat pipe with an electron-beam-welded or vacuum-brazed end cap.

PIV measurement at the blowdown pipe outlet

International Nuclear Information System (INIS)

Puustinen, M.; Laine, J.; Raesaenen, A.; Pyy, L.; Telkkae, J.

2013-04-01

This report summarizes the findings of the PIV measurement tests carried out in January - February 2013 with the scaled down PPOOLEX test facility at LUT. The main objective of the tests was to find out the operational limits of the PIV system regarding suitable test conditions and correct values of different adjustable PIV parameters. An additional objective was to gather CFD grade data for verification/validation of numerical models. Both water and steam injection tests were carried out. PIV measurements with cold water injection succeeded well. Raw images were of high quality, averaging over the whole measurement period could be done and flow fields close to the blowdown pipe outlet could be determined. In the warm water injection cases the obtained averaged velocity field images were harder to interpret, especially if the blowdown pipe was also filled with warm water in the beginning of the measurement period. The absolute values of the velocity vectors seemed to be smaller than in the cold water injection cases. With very small steam flow rates the steam/water interface was inside the blowdown pipe and quite stable in nature. The raw images were of good quality but due to some fluctuation in the velocity field averaging of the velocity images over the whole measured period couldn't be done. Condensation of steam in the vicinity of the pipe exit probably caused these fluctuations. A constant outflow was usually followed by a constant inflow towards the pipe exit. Vector field images corresponding to a certain phase of the test could be extracted and averaged but this would require a very careful analysis so that the images could be correctly categorized. With higher steam flow rates rapid condensation of large steam bubbles created small gas bubbles which were in front of the measurement area of the PIV system. They disturbed the measurements by reflecting laser light like seeding particles and therefore the raw images were of poor quality and they couldn't be

Residual stress improvement for pipe weld by means of induction heating pre-flawed pipe

International Nuclear Information System (INIS)

Umemoto, T.; Yoshida, K.; Okamoto, A.

1980-01-01

The intergranular stress corrosion cracking (IGSCC) has been found in type 304 stainless steel piping of several BWR plants. It is already well known that IGSCC is most likely to occur when three essential factors, material sensitization, high tensile stress and corrosive environment, are present. If the welding residual stress is sufficiently high (200 to approximately 400 MPa) in the inside piping surface near the welded joint, then it may be one of the biggest contributors to IGSCC. If the residual stress is reduced or reversed by some way, the IGSCC will be effectively mitigated. In this paper a method to improve the residual stress named IHSI (Induction Heating Stress Improvement) is explained. IHSI aims to improve the condition of residual stress in the inside pipe surface using the thermal stress induced by the temperature difference in pipe wall, that is produced when the pipe is heated from the outside surface by an induction heating coil and cooled on the inside surface by water simultaneously. This method becomes more attractive when it can be successfully applied to in-service piping which might have some pre-flaw. In order to verify the validity of IHSI for such piping, some experiments and calculations using finite element method were conducted. These results are mainly discussed in this paper from the view-points of residual stress, flaw behaviour during IHSI and material deterioration. (author)

A Novel Experimental Set-Up for Improving the Sensitivity of SV Waves to Shallow Surface-Breaking Cracks

Energy Technology Data Exchange (ETDEWEB)

Pecorari, Claudio [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Aeronautical and Vehicle Engineering

2006-03-15

Conventional inspection procedures to detect surface-breaking defects in train axels and thick pipes often employ 45-degree incidence shear vertical (SV) waves as probing tool. Recently obtained theoretical and experimental results indicate that this method is considerably less sensitivity to shallow surface-breaking defects, than the one in which the angle of incidence is selected to be close to the critical angle of the longitudinal wave. This project has confirmed this thesis by experimentally investigating the backscattering of SV waves by surface-breaking cracks as a function o t the angle of incidence. To this end, three cracks of depth approximately equal to 0.3 mm, 0.5 mm and 0.7 were introduced on the surface of steel samples with a thickness of 47 mm. These cracks were insonified with transducers operating at 2.25 MHz, 3.5 MHz, and 5 MHz, which correspond to wavelengths in steel of 1.38 mm, 0.88 mm, and 0.62 mm, respectively. The increase in sensitivity has been assessed in the order of 15 dB.

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)

Leachate storage transport tanker loadout piping

International Nuclear Information System (INIS)

Whitlock, R.W.

1994-01-01

This report shows the modifications to the W-025 Trench No. 31 leachate loadout discharge piping, and also the steps involved in installing the discharge piping, including dimensions and welding information. The installation of the discharge pipe should be done in accordance to current pipe installation standards. Trench No. 31 is a radioactive mixed waste land disposal facility

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Water hammer in elastic pipes

International Nuclear Information System (INIS)

Gale, J.; Tiselj, I.

2002-01-01

One dimensional two-fluid six-equation model of two-phase flow, that can be found in computer codes like RELAP5, TRAC, and CATHARE, was upgraded with additional terms, which enable modelling of the pressure waves in elastic pipes. It is known that pipe elasticity reduces the propagation velocity of the shock and other pressure waves in the piping systems. Equations that include the pipe elasticty terms are used in WAHA code, which is being developed within the WAHALoads project of 5't'h EU research program.(author)

Heat pipe dynamic behavior

Science.gov (United States)

Issacci, F.; Roche, G. L.; Klein, D. B.; Catton, I.

1988-01-01

The vapor flow in a heat pipe was mathematically modeled and the equations governing the transient behavior of the core were solved numerically. The modeled vapor flow is transient, axisymmetric (or two-dimensional) compressible viscous flow in a closed chamber. The two methods of solution are described. The more promising method failed (a mixed Galerkin finite difference method) whereas a more common finite difference method was successful. Preliminary results are presented showing that multi-dimensional flows need to be treated. A model of the liquid phase of a high temperature heat pipe was developed. The model is intended to be coupled to a vapor phase model for the complete solution of the heat pipe problem. The mathematical equations are formulated consistent with physical processes while allowing a computationally efficient solution. The model simulates time dependent characteristics of concern to the liquid phase including input phase change, output heat fluxes, liquid temperatures, container temperatures, liquid velocities, and liquid pressure. Preliminary results were obtained for two heat pipe startup cases. The heat pipe studied used lithium as the working fluid and an annular wick configuration. Recommendations for implementation based on the results obtained are presented. Experimental studies were initiated using a rectangular heat pipe. Both twin beam laser holography and laser Doppler anemometry were investigated. Preliminary experiments were completed and results are reported.

Detection and sizing of large-scale cracks in centrifugally cast stainless steel pipes using Lamb waves

International Nuclear Information System (INIS)

Ngoc, T.D.K.; Avioli, M.J. Jr.

1988-01-01

Application of conventional ultrasonic nondestructive evaluation (NDE) techniques to centrifugally cast stainless steel (CCSS) pipes in pressurized water reactors (PWRs) has been limited, mainly due to the anisotropy of the CCSS materials. Phenomena such as beam skewing and distortion are directly attributable to this anisotropy and cause severe difficulties in crack detection and sizing. To improve CCSS inspectability, the feasibility of using Lamb waves as the probing mechanism for detecting and characterizing a surface-breaking crack originating from the pipe interior surface is discussed. A similar research effort has been reported by Rokhlin who investigated the interaction of Lamb waves with delaminations in thin sheets. Rokhlin and Adler also reported recently on the use of Lamb waves for evaluating spot welds. The motivation for using this probing mechanism derives from the recognition that the difficulties introduced by beam skewing, beam distortion, and high attenuation are circumvented, since Lamb waves are not bulk waves, but are resonant vibrational modes of a solid plate

Anomalous dissipation and kinetic-energy distribution in pipes at very high Reynolds numbers.

Science.gov (United States)

Chen, Xi; Wei, Bo-Bo; Hussain, Fazle; She, Zhen-Su

2016-01-01

A symmetry-based theory is developed for the description of (streamwise) kinetic energy K in turbulent pipes at extremely high Reynolds numbers (Re's). The theory assumes a mesolayer with continual deformation of wall-attached eddies which introduce an anomalous dissipation, breaking the exact balance between production and dissipation. An outer peak of K is predicted above a critical Re of 10^{4}, in good agreement with experimental data. The theory offers an alternative explanation for the recently discovered logarithmic distribution of K. The concept of anomalous dissipation is further supported by a significant modification of the k-ω equation, yielding an accurate prediction of the entire K profile.

Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

International Nuclear Information System (INIS)

Lydell, B.

1999-05-01

This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research

Determination of the pipe stemming load

International Nuclear Information System (INIS)

Cowin, S.C.

1979-01-01

A mechanical model for the emplacement pipe system is developed. The model is then employed to determine the force applied to the surface collar of the emplacement pipe, the pipe-stemming load, and the stress along the emplacement pipe as a function of stemming height. These results are presented as integrals and a method for their numerical integration is given

Lightweight Heat Pipes Made from Magnesium

Science.gov (United States)

Rosenfeld, John N.; Zarembo, Sergei N.; Eastman, G. Yale

2010-01-01

Magnesium has shown promise as a lighter-weight alternative to the aluminum alloys now used to make the main structural components of axially grooved heat pipes that contain ammonia as the working fluid. Magnesium heat-pipe structures can be fabricated by conventional processes that include extrusion, machining, welding, and bending. The thermal performances of magnesium heat pipes are the same as those of equal-sized aluminum heat pipes. However, by virtue of the lower mass density of magnesium, the magnesium heat pipes weigh 35 percent less. Conceived for use aboard spacecraft, magnesium heat pipes could also be attractive as heat-transfer devices in terrestrial applications in which minimization of weight is sought: examples include radio-communication equipment and laptop computers.

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)

Heat pipe turbine vane cooling

Energy Technology Data Exchange (ETDEWEB)

Langston, L.; Faghri, A. [Univ. of Connecticut, Storrs, CT (United States)

1995-10-01

The applicability of using heat pipe principles to cool gas turbine vanes is addressed in this beginning program. This innovative concept involves fitting out the vane interior as a heat pipe and extending the vane into an adjacent heat sink, thus transferring the vane incident heat transfer through the heat pipe to heat sink. This design provides an extremely high heat transfer rate and an uniform temperature along the vane due to the internal change of phase of the heat pipe working fluid. Furthermore, this technology can also eliminate hot spots at the vane leading and trailing edges and increase the vane life by preventing thermal fatigue cracking. There is also the possibility of requiring no bleed air from the compressor, and therefore eliminating engine performance losses resulting from the diversion of compressor discharge air. Significant improvement in gas turbine performance can be achieved by using heat pipe technology in place of conventional air cooled vanes. A detailed numerical analysis of a heat pipe vane will be made and an experimental model will be designed in the first year of this new program.

Laboratory exercises on oscillation modes of pipes

Science.gov (United States)

Haeberli, Willy

2009-03-01

This paper describes an improved lab setup to study the vibrations of air columns in pipes. Features of the setup include transparent pipes which reveal the position of a movable microphone inside the pipe; excitation of pipe modes with a miniature microphone placed to allow access to the microphone stem for open, closed, or conical pipes; and sound insulation to avoid interference between different setups in a student lab. The suggested experiments on the modes of open, closed, and conical pipes, the transient response of a pipe, and the effect of pipe diameter are suitable for introductory physics laboratories, including laboratories for nonscience majors and music students, and for more advanced undergraduate laboratories. For honors students or for advanced laboratory exercises, the quantitative relation between the resonance width and damping time constant is of interest.

Regulatory aspects of the Leak Before Break application

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic Tuma, J.

2000-01-01

In the present paper the Leak Before Break (LBB) methodology is described on the way, as it was evaluated in Slovenian Nuclear Safety Administration in the process of the nuclear power plant (NPP) Krsko modernization process. In the recent decade, reviewed regulatory position regarding elimination the dynamic effects of the postulated primary coolant pipe ruptures in some countries was issued. The basis for such new approach are research achievements on different areas of science like metallurgy, fracture mechanics, dynamic analysis and testing of materials. By this new regulatory position the utility has the possibility to adopt LBB concept, but has to fulfill at least general prerequisites, described in standard review plan, where the basic principles and objectives of the evaluation process are described. World wide practice shows that more intensive and detailed evaluation is necessary as it is described in the standard review plan. The concerns and requests raising during evaluation of the consequences of the adopted LBB concept have generally some common points comparing regulatory experience of different countries which have already accepted this methodology or are currently under the process of evaluation. But nevertheless every nuclear power plant is unique, comparing specific material properties, dynamic analysis assumptions, safety analyses performed etc. One of the most important areas in the LBB evaluation process is reliable primary piping material evaluation. Several generic material studies were performed and the applicability of these studies for the justification of LBB has to be carefully performed. Elimination of the double ended guillotine break from the safety analysis has very strong impact to the safety analysis results and finally to the structures, systems and components too. Nuclear safety administration has made a lot of effort to make its position on this methodology clear. The expertise of the authorized institutions is in such process