WorldWideScience

Sample records for pipe break testing

  1. 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.

  2. 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

  3. Leak before break piping evaluation diagram

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  4. 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.).

  5. View of industry on the impact of pipe break criteria

    International Nuclear Information System (INIS)

    Bernsen, S.A.

    1983-01-01

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

  6. Statistical models for the analysis of water distribution system pipe break data

    International Nuclear Information System (INIS)

    Yamijala, Shridhar; Guikema, Seth D.; Brumbelow, Kelly

    2009-01-01

    The deterioration of pipes leading to pipe breaks and leaks in urban water distribution systems is of concern to water utilities throughout the world. Pipe breaks and leaks may result in reduction in the water-carrying capacity of the pipes and contamination of water in the distribution systems. Water utilities incur large expenses in the replacement and rehabilitation of water mains, making it critical to evaluate the current and future condition of the system for maintenance decision-making. This paper compares different statistical regression models proposed in the literature for estimating the reliability of pipes in a water distribution system on the basis of short time histories. The goals of these models are to estimate the likelihood of pipe breaks in the future and determine the parameters that most affect the likelihood of pipe breaks. The data set used for the analysis comes from a major US city, and these data include approximately 85,000 pipe segments with nearly 2500 breaks from 2000 through 2005. The results show that the set of statistical models previously proposed for this problem do not provide good estimates with the test data set. However, logistic generalized linear models do provide good estimates of pipe reliability and can be useful for water utilities in planning pipe inspection and maintenance

  7. 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)

  8. 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)

  9. Pipe break prediction based on evolutionary data-driven methods with brief recorded data

    International Nuclear Information System (INIS)

    Xu Qiang; Chen Qiuwen; Li Weifeng; Ma Jinfeng

    2011-01-01

    Pipe breaks often occur in water distribution networks, imposing great pressure on utility managers to secure stable water supply. However, pipe breaks are hard to detect by the conventional method. It is therefore necessary to develop reliable and robust pipe break models to assess the pipe's probability to fail and then to optimize the pipe break detection scheme. In the absence of deterministic physical models for pipe break, data-driven techniques provide a promising approach to investigate the principles underlying pipe break. In this paper, two data-driven techniques, namely Genetic Programming (GP) and Evolutionary Polynomial Regression (EPR) are applied to develop pipe break models for the water distribution system of Beijing City. The comparison with the recorded pipe break data from 1987 to 2005 showed that the models have great capability to obtain reliable predictions. The models can be used to prioritize pipes for break inspection and then improve detection efficiency.

  10. Sensitivity Study of the Peak Cladding Temperature for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    International Nuclear Information System (INIS)

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

    2005-12-01

    The effect of the thermal hydraulic operation parameters, the stroke times of safety-related valves, the node number of test fuel for MARS modeling, and the axial power distribution on the peak cladding temperature (PCT) has been investigated for the loss of coolant accident of the 3-pin fuel test loop. The thermal hydraulic operation parameters investigated are the thermal power of the fuel test loop and the flow rate, temperature, and pressure of the main cooling water. The effect of the thermal power and the coolant temperature on the peak cladding temperature is dominant as compared with that of the coolant flow rate and pressure. The maximum PCT increases up to about 34.3K for the room 1 LOCA when the thermal power increase by 5% of the normal operation power and decreases up to about 38.9K for the room 1 LOCA when the coolant temperature decrease by 2% of the normal operation temperature. The effect of the stroke time of the loop isolation valves on the PCT is also dominant. However the effect of the stroke time of the safety injection valves and depressurization vent valves are negligible. Especially the maximum PCT increases up to 25.7K with the increase of the design stroke time of the cold leg loop isolation valve by 13% and decreases up to 25.1K with the decrease of the design stroke time by 13%. The maximum PCT increases by 3.3K as the number of nodes increases from 7 to 14 for the MARS model of test fuel. Three different axial power distributions are also investigated. The maximum PCT occurs for the room 1 LOCA in case the peak power is shifted to the downstream by 20cm

  11. 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)

  12. 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).

  13. 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)

  14. 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)

  15. 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.

  16. 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

  17. 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.

  18. 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)

  19. Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

    International Nuclear Information System (INIS)

    Wendel, M.W.; Chen, N.C.J.; Yoder, G.L.

    1994-01-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

  20. 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.)

  1. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 3. Evaluation of potential for pipe breaks

    Energy Technology Data Exchange (ETDEWEB)

    1984-11-01

    The Executive Director for Operations (EDO) in establishing the Piping Review Committee concurred in its overall scope that included an evaluation of the potential for pipe breaks. The Pipe Break Task Group has responded to this directive. This report summarizes a review of regulatory documents and contains the Task Group's recommendations for application of the leak-before-break (LBB) approach to the NRC licensing process. The LBB approach means the application of fracture mechanics technology to demonstrate that high energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent as longitudinal or diagonal splits. The Task Group's reommendations and discussion are founded on current and ongoing NRC staff actions as presented in Section 3.0 of this report. Additional more detailed comments and discussion are presented in Section 5.0 and in Appendices A and B. The obvious issues are the reexamination of the large pipe break criteria and the implications of any changes in the criteria as they influence items such as jet loads and pipe whip. The issues have been considered and the Task Group makes the following recommendations.

  2. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 3. Evaluation of potential for pipe breaks

    International Nuclear Information System (INIS)

    1984-11-01

    The Executive Director for Operations (EDO) in establishing the Piping Review Committee concurred in its overall scope that included an evaluation of the potential for pipe breaks. The Pipe Break Task Group has responded to this directive. This report summarizes a review of regulatory documents and contains the Task Group's recommendations for application of the leak-before-break (LBB) approach to the NRC licensing process. The LBB approach means the application of fracture mechanics technology to demonstrate that high energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent as longitudinal or diagonal splits. The Task Group's reommendations and discussion are founded on current and ongoing NRC staff actions as presented in Section 3.0 of this report. Additional more detailed comments and discussion are presented in Section 5.0 and in Appendices A and B. The obvious issues are the reexamination of the large pipe break criteria and the implications of any changes in the criteria as they influence items such as jet loads and pipe whip. The issues have been considered and the Task Group makes the following recommendations

  3. 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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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 ...

  10. 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

  11. Bayesian Belief Networks for predicting drinking water distribution system pipe breaks

    International Nuclear Information System (INIS)

    Francis, Royce A.; Guikema, Seth D.; Henneman, Lucas

    2014-01-01

    In this paper, we use Bayesian Belief Networks (BBNs) to construct a knowledge model for pipe breaks in a water zone. To the authors’ knowledge, this is the first attempt to model drinking water distribution system pipe breaks using BBNs. Development of expert systems such as BBNs for analyzing drinking water distribution system data is not only important for pipe break prediction, but is also a first step in preventing water loss and water quality deterioration through the application of machine learning techniques to facilitate data-based distribution system monitoring and asset management. Due to the difficulties in collecting, preparing, and managing drinking water distribution system data, most pipe break models can be classified as “statistical–physical” or “hypothesis-generating.” We develop the BBN with the hope of contributing to the “hypothesis-generating” class of models, while demonstrating the possibility that BBNs might also be used as “statistical–physical” models. Our model is learned from pipe breaks and covariate data from a mid-Atlantic United States (U.S.) drinking water distribution system network. BBN models are learned using a constraint-based method, a score-based method, and a hybrid method. Model evaluation is based on log-likelihood scoring. Sensitivity analysis using mutual information criterion is also reported. While our results indicate general agreement with prior results reported in pipe break modeling studies, they also suggest that it may be difficult to select among model alternatives. This model uncertainty may mean that more research is needed for understanding whether additional pipe break risk factors beyond age, break history, pipe material, and pipe diameter might be important for asset management planning. - Highlights: • We show Bayesian Networks for predictive and diagnostic management of water distribution systems. • Our model may enable system operators and managers to prioritize system

  12. 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)

  13. 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.

  14. 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

  15. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  16. Structural Break Tests Robust to Regression Misspecification

    Directory of Open Access Journals (Sweden)

    Alaa Abi Morshed

    2018-05-01

    Full Text Available Structural break tests for regression models are sensitive to model misspecification. We show—analytically and through simulations—that the sup Wald test for breaks in the conditional mean and variance of a time series process exhibits severe size distortions when the conditional mean dynamics are misspecified. We also show that the sup Wald test for breaks in the unconditional mean and variance does not have the same size distortions, yet benefits from similar power to its conditional counterpart in correctly specified models. Hence, we propose using it as an alternative and complementary test for breaks. We apply the unconditional and conditional mean and variance tests to three US series: unemployment, industrial production growth and interest rates. Both the unconditional and the conditional mean tests detect a break in the mean of interest rates. However, for the other two series, the unconditional mean test does not detect a break, while the conditional mean tests based on dynamic regression models occasionally detect a break, with the implied break-point estimator varying across different dynamic specifications. For all series, the unconditional variance does not detect a break while most tests for the conditional variance do detect a break which also varies across specifications.

  17. Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes Savannah River Project L and P Reactors

    International Nuclear Information System (INIS)

    Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P.

    1989-01-01

    A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. the potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break

  18. 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.

  19. 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)

  20. 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)

  1. 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)

  2. ANSPipe: An IBM-PC interactive code for pipe-break assessment

    International Nuclear Information System (INIS)

    Fullwood, R.R.; Harrington, M.

    1988-01-01

    The advanced neutron source (ANS) being designed at Oak Ridge National Laboratory will be the world's highest flux neutron source and best facility for associated basic and applied research. The ANSPipe code was written as an aid for the piping configuration and material selection to enhance safety and availability. The primary calculation is based on the Thomas mode. which models pipe leak or break probabilities as proportional to the length of the segment and diameter and the inverse square of the wall thickness. This scaling, based on experience, is adjusted for radiation effects, using the Regulatory Guide 1.99 model, and for cyclic fatigue, stress corrosion, and inspection, using adaptations form the PRAISE-B code. The key to an ANSPipe analysis is the definition of the pipe segments. A pipe segment is defined as a length of pipe in which all the parameters affecting the pipe are constant or reasonably so. Thus, a segment would be a length of pipe of constant diameter, thickness, material type, internal pressure, flux distribution, stress, and submergence or nonsubmergence

  3. Performance testing of a hydrogen heat pipe

    International Nuclear Information System (INIS)

    Alario, J.; Kosson, R.

    1980-01-01

    Test results are presented for a reentrant groove heat pipe with hydrogen working fluid. The heat pipe became operational between 20 and 30 K after a cooldown from 77 K without any difficulty. Steady-state performance data taken over a 19 to 23 K temperature range indicated the following: (1) maximum heat transport capacity 5.4 W-m (2) static wicking height 1.42 cm and (3) overall heat pipe conductance 1.7 W/C. These data agreed remarkably well with extrapolations made from comparable ammonia test results. The maximum heat transport capacity is 9.5% larger than the extrapolated value, but the static wicking height is the same. The overall conductance is 29% of the ammonia value, which is close to the ratio of liquid thermal conductivities (24%). Also, recovery from a completely frozen condition was accomplished within 5 min by simply applying an evaporater heat load of 1.8 W

  4. 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.)

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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

  10. The concepts of leak before break and absolute reliability of NPP equipment and piping

    International Nuclear Information System (INIS)

    Getman, A.F.; Komarov, O.V.; Sokov, L.M.

    1997-01-01

    This paper describes the absolute reliability (AR) concept for ensuring safe operation of nuclear plant equipment and piping. The AR of a pipeline or component is defined as the level of reliability when the probability of an instantaneous double-ended break is near zero. AR analysis has been applied to Russian RBMK and VVER type reactors. It is proposed that analyses required for application of the leak before break concept should be included in AR implementation. The basic principles, methods, and approaches that provide the basis for implementing the AR concept are described

  11. The concepts of leak before break and absolute reliability of NPP equipment and piping

    Energy Technology Data Exchange (ETDEWEB)

    Getman, A.F.; Komarov, O.V.; Sokov, L.M. [and others

    1997-04-01

    This paper describes the absolute reliability (AR) concept for ensuring safe operation of nuclear plant equipment and piping. The AR of a pipeline or component is defined as the level of reliability when the probability of an instantaneous double-ended break is near zero. AR analysis has been applied to Russian RBMK and VVER type reactors. It is proposed that analyses required for application of the leak before break concept should be included in AR implementation. The basic principles, methods, and approaches that provide the basis for implementing the AR concept are described.

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. Seismic testing and analysis of a prototypic nonlinear piping system

    International Nuclear Information System (INIS)

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

    1982-11-01

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

  5. Internal testing of pipe systems with IRIS inspection system

    International Nuclear Information System (INIS)

    1986-01-01

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

  6. 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)

  7. 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.

  8. The IPIRG-1 pipe system fracture tests: Experimental results

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  9. 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.)

  10. Small-bore-piping seismic-test findings

    International Nuclear Information System (INIS)

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

    1981-12-01

    Description is given of a test series in which a 1-inch diameter stainless steel pipe system was subjected to dynamic testing. The test system consisted of approximately 40-feet of schedule 40 pipe, with several bends and risers, supported from a rigid test frame. FFTF prototypic pipe clamps, dead weight supports, mechanical snubbers, and insulation were utilized. Several variations of the pipe support configuration were tested. Measured test results are compared with analytical predictions for each configuration. Plans for future testing are discussed

  11. Fatigue evaluation of piping systems with limited vibration test data

    International Nuclear Information System (INIS)

    Huang, S.N.

    1990-11-01

    The safety-related piping in a nuclear power plant may be subjected to pump- or fluid-induced vibrations that, in general, affect only local areas of the piping systems. Pump- or fluid-induced vibrations typically are characterized by low levels of amplitudes and a high number of cycles over the lifetime of plant operation. Thus, the resulting fatigue damage to the piping systems could be an important safety concern. In general, tests and/or analyses are used to evaluate and qualify the piping systems. Test data, however, may be limited because of lack of instrumentation in critical piping locations and/or because of difficulty in obtaining data in inaccessible areas. This paper describes and summarizes a method to use limited pipe vibration test data, along with analytical harmonic response results from finite-element analyses, to assess the fatigue damage of nuclear power plant safety-related piping systems. 5 refs., 2 figs., 11 tabs

  12. 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)

  13. 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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Seismic test of high temperature piping for HTGR

    International Nuclear Information System (INIS)

    Kobatake, Kiyokazu; Midoriyama, Shigeru; Ooka, Yuzi; Suzuki, Michiaki; Katsuki, Taketsugu

    1983-01-01

    Since the high temperature pipings for the high temperature gas-cooled reactor contain helium gas at 1000 deg C and 40 kgf/cm 2 , the double-walled pipe type consisting of the external pipe serving as the pressure boundary and the internal pipe with heat insulating structure was adopted. Accordingly, their aseismatic design is one of the important subjects. Recently, for the purpose of grasping the vibration characteristics of these high temperature pipings and obtaining the data required for the aseismatic design, two specimens, that is, a double-walled pipe model and a heat-insulating structure, were made, and the vibration test was carried out on them, using a 30 ton vibration table of Kawasaki Heavy Industries Ltd. In the high temperature pipings of the primary cooling system for the multi-purpose, high temperature gas-cooled experimental reactor, the external pipes of 32 B bore as the pressure boundary and the internal pipes of 26 B bore with internal heat insulation consisting of double layers of fiber and laminated metal insulators as the temperature boundary were adopted. The testing method and the results are reported. As the spring constant of spacers is larger and clearance is smaller, the earthquake wave response of double-walled pipes is smaller, and it is more advantageous. The aseismatic property of the heat insulation structure is sufficient. (Kako, I.)

  16. 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.

  17. 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

  18. 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)

  19. 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.)

  20. Finite-element analysis of flawed and unflawed pipe tests

    International Nuclear Information System (INIS)

    James, R.J.; Nickell, R.E.; Sullaway, M.F.

    1989-12-01

    Contemporary versions of the general purpose, nonlinear finite element program ABAQUS have been used in structural response verification exercises on flawed and unflawed austenitic stainless steel and ferritic steel piping. Among the topics examined, through comparison between ABAQUS calculations and test results, were: (1) the effect of using variations in the stress-strain relationship from the test article material on the calculated response; (2) the convergence properties of various finite element representations of the pipe geometry, using shell, beam and continuum models; (3) the effect of test system compliance; and (4) the validity of ABAQUS J-integral routines for flawed pipe evaluations. The study was culminated by the development and demonstration of a ''macroelement'' representation for the flawed pipe section. The macroelement can be inserted into an existing piping system model, in order to accurately treat the crack-opening and crack-closing static and dynamic response. 11 refs., 20 figs., 1 tab

  1. Leak test of the pipe line for radioactive liquid waste

    International Nuclear Information System (INIS)

    Machida, Chuji; Mori, Shoji.

    1976-01-01

    In the Tokai Research Establishment, most of the radioactive liquid waste is transferred to a wastes treatment facility through pipe lines. As part of the pipe lines a cast iron pipe for town gas is used. Leak test has been performed on all joints of the lines. For the joints buried underground, the test was made by radioactivity measurement of the soil; and for the joints in drainage ditch by the pressure and bubble methods. There were no leakage at all, indicating integrity of all the joints. On the other hand, it is also known by the other test that the corrosion of inner surface of the piping due to liquid waste is only slight. The pipe lines for transferring radioactive liquid waste are thus still usable. (auth.)

  2. Significance of high level test data in piping design

    International Nuclear Information System (INIS)

    McLean, J.L.; Bitner, J.L.

    1991-01-01

    During the 1980's the piping technical community in the U.S. initiated a series of research activities aimed at reducing the conservatism inherent in nuclear piping design. One of these activities was directed at the application of the ASME Code rules to the design of piping subjected to dynamic loads. This paper surveys the test data obtained from three groups in the U.S. and none in the U.K., and correlates the findings as they relate to the failure modes of piping subjected to seismic loads. The failure modes experienced as the result of testing at dynamic loads significantly in excess of anticipated loads specified for any of the ASME Code service levels are discussed. A recommendation is presented for modifying the Code piping rules to reduce the conservatism inherent in seismic design

  3. 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.

  4. 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.

  5. 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)

  6. Inelastic analysis of Battelle-Columbus piping elbow creep test

    International Nuclear Information System (INIS)

    Dhalla, A.K.; Newman, S.Z.

    1979-01-01

    Analytical results are presented for room temperature and 593 deg. C creep bending deformation of a piping elbow structure tested at the Battelle-Columbus Laboratory. This analysis was performed in support of the International Piping Benchmark Problem Program being coordinated by ORNL. Results are presented for both simplified and refined structural models, and compared with test measurements reported by the Battelle-Columbus Laboratory. (author)

  7. 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

  8. 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.

  9. Methodology for Life Testing of Refractory Metal / Sodium Heat Pipes

    International Nuclear Information System (INIS)

    Martin, James J.; Reid, Robert S.

    2006-01-01

    This work establishes an approach to generate carefully controlled data to find heat pipe operating life with material-fluid combinations capable of extended operation. To accomplish this goal acceleration is required to compress 10 years of operational life into 3 years of laboratory testing through a combination of increased temperature and mass fluence. Specific test series have been identified, based on American Society for Testing and Materials (ASTM) specifications, to investigate long-term corrosion rates. The refractory metal selected for demonstration purposes is a molybdenum-44.5% rhenium alloy formed by powder metallurgy. The heat pipes each have an annular crescent wick formed by hot isostatic pressing of molybdenum-rhenium wire mesh. The heat pipes are filled by vacuum distillation with purity sampling of the completed assembly. Round-the-clock heat pipe tests with 6-month destructive and non-destructive inspection intervals are conducted to identify the onset and level of corrosion. Non-contact techniques are employed to provide power to the evaporator (radio frequency induction heating at 1 to 5 kW per heat pipe) and calorimetry at the condenser (static gas gap coupled water cooled calorimeter). The planned operating temperature range extends from 1123 to 1323 K. Accomplishments before project cancellation included successful development of the heat pipe wick fabrication technique, establishment of all engineering designs, baseline operational test requirements, and procurement/assembly of supporting test hardware systems. (authors)

  10. 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

  11. Structural damping results from vibration tests of straight piping sections

    International Nuclear Information System (INIS)

    Ware, A.G.; Thinnes, G.L.

    1984-01-01

    EG and G Idaho is assisting the USNRC and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on 76-mm and 203-mm (3-in. amd 8-in.) Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 10-m (33-ft) straight sections of piping were rigidly supported at the ends. Spring, rod, and constant force hangers, as well as a sway brace and snubbers were included as intermediate supports. Excitation was provided by low-force level hammer inpacts, a hydraulic shaker, and a 445-kN (50-ton) overhead crane. Data was recorded using acceleration, strain, and displacement time histories. This paper presents results from the testing showing the effect of stress level and type of supports on structural damping in piping

  12. Seismic fragility test of a 6-inch diameter pipe system

    International Nuclear Information System (INIS)

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

    1987-02-01

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

  13. Corrosion evaluation in insulated pipes by non destructive testing method

    International Nuclear Information System (INIS)

    Abd Razak Hamzah; Azali Muhammad; Mohammad Pauzi Ismail; Abd Nassir Ibrahim; Abd Aziz Mohamed; Sufian Saad; Saharuddin Sayuti; Shukri Ahmad

    2002-01-01

    In engineering plants, detection of corrosion and evaluation of deposit in insulated pipes using radiography method are considered as a very challenging tasks. In General this degradation problem is attributed to water condensation. It causes the formation of deposit and scale inside the pipe, as well as between the insulation and pipe in cold temperature pipes. On the other hand, for hot temperature pipes the main problem is mainly due to corrosion/erosion attack inside the pipe. In the study of corrosion in pipelines, one of the most important parameters to be monitored and measured is the wall thickness. Currently, most pipeline corrosion monitoring and evaluation for both insulated and non-insulated pipes is performed using an ultrasonic method. The most common technique is that based on the A-Scan, using either a normal flaw detector or some form of dedicated equipment. However, with recent development of ultrasonic technology, more advance method, namely B-Scan and C-scan techniques are also available. The most notable disadvantage of using this method is that the insulation covering the pipe has to be removed before the inspection can be carried out and this is considered as not so cost effective. Due to this reason, the possibility of employing other alternative NDT method, namely radiographic testing method were studied. The technique used in this studied are known as tangential technique. In this study it was found that the result found using tangential technique is consistent with the actual thickness of the pipe. Result of this study is presented and discussed in this paper. (Author)

  14. Development and testing of restraints for nuclear piping systems

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  15. In-Pipe Wireless Communication for Underground Sampling and Testing

    NARCIS (Netherlands)

    Nguyen, Nhan D.T.; Le, Duc V.; Meratnia, Nirvana; Havinga, Paul J.M.

    2017-01-01

    In this paper, we present an effective and low- cost wireless communication system for extremely long and narrow pipes that can replay the extant wire system in underground sensor network applications such as soil sampling and testing with the Cone Penetration Test (CPT), the most widely used

  16. Bayesian analysis of heat pipe life test data for reliability demonstration testing

    International Nuclear Information System (INIS)

    Bartholomew, R.J.; Martz, H.F.

    1985-01-01

    The demonstration testing duration requirements to establish a quantitative measure of assurance of expected lifetime for heat pipes was determined. The heat pipes are candidate devices for transporting heat generated in a nuclear reactor core to thermoelectric converters for use as a space-based electric power plant. A Bayesian analysis technique is employed, utilizing a limited Delphi survey, and a geometric mean accelerated test criterion involving heat pipe power (P) and temperature (T). Resulting calculations indicate considerable test savings can be achieved by employing the method, but development testing to determine heat pipe failure mechanisms should not be circumvented

  17. 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)

  18. 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)

  19. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Anglaret, G.; Lasne, M.

    1983-08-01

    Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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.)

  6. 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.)

  7. 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.)

  8. Specifying and manufacturing piping for the fast flux test facility

    International Nuclear Information System (INIS)

    Moen, R.A.; O'Keefe, D.P.; Irvin, J.E.; Tobin, J.C.

    1974-01-01

    Specification of materials for liquid metal reactor coolant piping, at service temperatures up to 1200 0 F, involves a number of considerations unique to these systems. The mechanical property/design allowable stress considerations which led to the selection and specification of specific materials for the Fast Flux Test Facility piping are discussed. Additional considerations are described indicating allowances made for material changes anticipated in service. These measures primarily involved raising the minimum carbon content to a value that would insure the strength of the material always remains above that assumed in the initial design, although other considerations are discussed. The processes by which this piping was manufactured, its resulting characteristics and methods of subsequent handling/assembly are briefly discussed. (U.S.)

  9. Test method for measuring insulation values of cryogenic pipes

    NARCIS (Netherlands)

    Velthuis, J.F.M.; Blokland, H.; Klaver, B.W.; Beld, C. van de

    2010-01-01

    In this paper a large-area heat flux and temperature sensor (HFT) is used for the evaluation of the insulation value of cryogenic pipes. The HFT is flexible and clamp-on. The test method is relatively simple and can be used in-situ. The HFT makes it possible to monitor insulation performance over

  10. 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)

  11. 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

  12. 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.

  13. 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

  14. 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

  15. Test of Seal System for Flexible Pipe End Fitting

    DEFF Research Database (Denmark)

    Banke, Lars; Jensen, Thomas Gregers

    1999-01-01

    The purpose of the end fitting seal system is to ensure leak proof termination of flexible pipes. The seal system of an NKT end fitting normally consists of a number of ring joint gaskets mounted in a steel sleeve on the outside of the polymeric inner liner of the pipe. The seal system is activated...... by compression of the gaskets, thus using the geometry to establish a seal towards the inner liner of the pipe and the steel sleeve of the end fitting. This paper describes how the seal system of an end fitting can be tested using an autoclave. By regulating temperature and pressure, the seal system can...... be tested up to 130oC and 51.7 MPa. Pressure, temperature and the mechanical behaviours of the pipe are measured for use in further research. The set-up is used to test the efficiency of the seal system as function of parameters such as cross sectional shapes of the gaskets, tolerances between gaskets...

  16. 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%

  17. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  18. 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.

  19. 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.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  1. Niobium 1 percent zirconium/potassium and titanium/potassium life-test heat pipe design and testing

    Science.gov (United States)

    Sena, J. Tom; Merrigan, Michael A.

    Experimental lifetime performance studies currently in progress use Niobium 1 percent Zirconium (Nb-1Zr) and Titanium (Ti) heat pipes with potassium (K) as the working fluid. A heat pipe life test matrix was developed for testing the heat pipes. Because the corrosion rates in alkali metal heat pipes are affected by temperature and working fluid evaporation flux, the variable parameters of the experimental matrix are established as steady operating temperature and input heat flux density. Total impurity inventory is a factor in corrosion rate so impurity levels are being evaluated in the heat pipe materials before and after testing. Eight Nb-1Zr/K heat pipes were designed, fabricated, and tested. Two of the heat pipes have completed testing whereas the other six are currently in test. These are gravity assist heat pipes operating in a reflux mode. The heat pipes are tested by sets, one set of two and two sets of three heat pipes. Three Ti/K heat pipes are also in life test. These heat pipes are tested as a set in a horizontal position in a capillary pumped annular flow mode. Each of the heat pipes is encapsulated in a quartz vacuum container with a water calorimeter over the vacuum container for power throughput measurements. Thermocouples are attached to the heat pipes for measuring temperature. Heat input to the heat pipes is via an RF coil. The heat pipes are operating at between 800 and 900 K, with heat input fluxes of 13.8 to 30 W/sq cm. Of the Nb-1Zr/K heat pipes, two of the heat pipes have been in operation for 14,000 hours, three over 10,000 hours, and three over 7,000 hours. The Ti/K heat pipes have been in operation for 1,266 hours.

  2. ROSA-III 50 % break integral test RUN 916

    International Nuclear Information System (INIS)

    Yonomoto, Taisuke; Tasaka, Kanji; Koizumi, Yasuo; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Suzuki, Mitsuhiro; Murata, Hideo

    1985-08-01

    This report presents the experimental data of RUN 916 conducted at the ROSA-III test facility. The facility is a volumetrically scaled (1/424) simulator for a BWR/6 with the electrically heated core, the break simulator and the scaled ECCS(emergency core cooling system). RUN 916 was a 50 % split break test at the recirculation pump suction line with an assumption of HPCS diegel generator failure and conducted as one of the break area parameter tests. A peak cladding temperature (PCT) of 917 K was reached at 190 s after the break during the reflooding phase. Whole core was completely quenched by ECCS, and the effectiveness of ECCS was confermed. The primary test results of RUN 916 are compared in this report with those of RUN 926, which was a 200 % double-ended break test. The initiation of core dryout in RUN 916 was later than that in RUN 926 because of the smaller discharge flow rate. Duration of core dryourt was, however, longer in RUN 916 because of later actuation of ECCSs. PCT in RUN 916 was 133 K higher than that in RUN 926. (author)

  3. ROSA-III 100 % break integral test Run 914

    International Nuclear Information System (INIS)

    Yonomoto, Taisuke; Tasaka, Kanji; Koizumi, Yasuo; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Suzuki, Mitsuhiro; Murata, Hideo

    1987-05-01

    This report presents the experimental data of RUN 914 conducted at the ROSA-III test facility. The facility is a volumetrically scaled (1/424) simulator for a BWR/6 with the electrically heated core, the break simulator and the scaled ECCS (emergency core cooling system). RUN 914 was a 100% split break test at the recirculation pump suction line with an assumption of HPCS diesel generator failure and conducted as one of the break area parameter tests. A peak cladding temperature (PCT) of 851 K was reached at 130 s after the break during the reflooding phase. Whole core was completely quenched by ECCS, and the effectiveness of ECCS was confirmed. The primary test results of RUN 914 are compared in this report with those of RUN 926, which was a 200 % double-ended break test. The initiation of core dryout in RUN 914 was almost the same as that in RUN 926. Duration of core dryourt was, however, longer in RUN 914 because of later actuation of ECCSs. PCT in RUN 914 was 67 K higher than that in RUN 926. (author)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. A procedure for estimation of pipe break probabilities due to IGSCC

    International Nuclear Information System (INIS)

    Bergman, M.; Brickstad, B.; Nilsson, F.

    1998-06-01

    A procedure has been developed for estimation of the failure probability of welds joints in nuclear piping susceptible to intergranular stress corrosion cracking. The procedure aims at a robust and rapid estimate of the failure probability for a specific weld with known stress state. Random properties are taken into account of the crack initiation rate, the initial crack length, the in-service inspection efficiency and the leak rate. A computer realization of the procedure has been developed for user friendly applications by design engineers. Some examples are considered to investigate the sensitivity of the failure probability to different input quantities. (au)

  9. Assessment of Pipe Wall Loss Using Guided Wave Testing

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. 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

  11. Testing the efficiency of the wine market using unit root tests with sharp and smooth breaks

    Directory of Open Access Journals (Sweden)

    Elie Bouri

    2017-12-01

    Full Text Available This paper examines the efficient market hypothesis for the wine market using a novel unit root test while accounting for sharp shifts and smooth breaks in the monthly data. We find evidence of structural shifts and nonlinearity in the wine indices. Contrary to the results from conventional linear unit root tests, when we account for sharp shifts and smooth breaks, the unit root null for each of the wine indices has been rejected. Overall, our results suggest that the wine market is inefficient when we incorporate breaks. We provide some practical and policy implications of our findings. Keywords: Wine market, Efficiency, Sharp and smooth breaks, Unit root tests

  12. Breaking diffeomorphism invariance and tests for the emergence of gravity

    International Nuclear Information System (INIS)

    Anber, Mohamed M.; Aydemir, Ufuk; Donoghue, John F.

    2010-01-01

    If general relativity is an emergent phenomenon, there may be small violations of diffeomorphism invariance. We propose a phenomenology of perturbatively small violations of general relativity by the inclusion of terms which break general covariance. These can be tested by matching to the parameterized post-Newtonian formalism. The most sensitive tests involve pulsar timing and provide an extremely strong bound, with a dimensionless constraint of order 10 -20 relative to gravitational strength.

  13. Software test attacks to break mobile and embedded devices

    CERN Document Server

    Hagar, Jon Duncan

    2013-01-01

    Address Errors before Users Find Them Using a mix-and-match approach, Software Test Attacks to Break Mobile and Embedded Devices presents an attack basis for testing mobile and embedded systems. Designed for testers working in the ever-expanding world of ""smart"" devices driven by software, the book focuses on attack-based testing that can be used by individuals and teams. The numerous test attacks show you when a software product does not work (i.e., has bugs) and provide you with information about the software product under test. The book guides you step by step starting with the basics. It

  14. Pipe Overpack Container Fire Testing: Phase I & II

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, Victor G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lopez, Carlos [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a new series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016, and described herein, were done in two phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. This report will describe the various tests conducted in phase I and II, present preliminary results from these tests, and discuss implications for the POCs.

  15. Pipe Overpack Container Fire Testing: Phase I II & III.

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, Victor G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lopez, Carlos [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-02-01

    The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016 were done in three phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. The goal of the third phase was to see if surrogate aerosol gets released from the PC when the drum lid is off. This report will describe the various tests conducted in phase I, II, and III, present preliminary results from these tests, and discuss implications for the POCs.

  16. 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.))

  17. 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)

  18. Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

    International Nuclear Information System (INIS)

    French, R.T.

    1975-08-01

    Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

  19. Fatigue test results of straight pipe with flaws in inner surface

    International Nuclear Information System (INIS)

    Shibata, Katsuyuki; Oba, Toshihiro; Kawamura, Takaichi; Yokoyama, Norio; Miyazono, Shohachiro

    1981-01-01

    Fatigue and fracture tests of piping models with flaws in the inner surface were carried out to investigate the fatigue crack growth, coalescence of multiple cracks and fracture behavior. Two straight test pipes with and without weldment in the test section of SUS304L stainless steel were tested under almost the same test conditions. Three artificial defects were machined in the inner surface of the test section of the test pipes. The fatigue test were performed untill the cracks coalesced and grew through the thickness. Subsequently, a static load was imposed on test pipe which contained a large crack in the test section. The test results show that the fatigue crack growth is slower than that predicted by the method specified in the Section XI of ASME Boiler and Pressure Vessel Code, and that the test pipes can endure more than the static load of 3Sm without an unstable fracture. (author)

  20. Seismic response and damping tests of small bore LMFBR piping and supports

    International Nuclear Information System (INIS)

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

    1984-01-01

    Seismic testing and analysis of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described. Measured responses to simulated seismic excitations are compared with analytical predictions based on NRC Regulatory Guide 1.61 and measured system damping values. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps

  1. An SBLOCA Test for Shutdown Cooling Line Break Using the SMART-ITL Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Kim, Dong Eok; Ryu, Sung Uk; Shin, Yong Cheol; Ko, Yung Joo; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The objectives of SMART-ITL are to investigate and understand the integral performance of the reactor systems and components, and the thermalhydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. Its height was preserved and its area and volume were scaled down to 1/49 compared with the SMART prototype plant. The SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The SMART was installed at KAERI and several transient tests were recently finished. In this paper, the test results for a steady-state operation and a transient of the small break loss of coolant accident (SBLOCA) are discussed. An SBLOCA test simulating the shutdown cooling line break was performed using SMART-ITL properly. All parameters were in good agreement with the target values during the steady-state operation period. The pressures and temperatures show reasonable behaviors during the SBLOCA test. SMART (System-integrated Modular Advanced ReacTor) which was designed by KAERI is an integral type reactor. The standard design approval for the SMART design was issued on July 4th of 2012 by a Korean regulatory body, the Nuclear Safety and Security Commission (NSSC). The main components including a pressurizer, steam generators, and reactor coolant pumps are installed in a reactor pressure vessel, and there are no large-size pipes. The safety systems could be simplified as an LBLOCA (Large-Break Loss of Coolant Accident) scenario is inherently excluded. An integral-effect test loop for SMART (SMART-ITL, or FESTA) was designed to simulate the integral thermal-hydraulic behavior of SMART. The SMART-ITL has been designed using a volume scaling methodology.

  2. Analysis of Semiscale Mod-1 integral test with asymmetrical break (Test S-29-1)

    International Nuclear Information System (INIS)

    Langerman, M.A.

    1977-03-01

    Selected experimental data obtained from Semiscale Mod-1 cold leg break Test S-29-1 and results obtained from analytical codes are analyzed. This test was the first integral blowdown reflood test conducted with the Mod-1 system and was a special test designed specifically to evaluate the sensitivity of the early Mod-1 core thermal response (0 to 5 sec after rupture) to the magnitude and direction of the core flow. To achieve this specific objective in Test S-29-1, the vessel side break area was reduced to approximately one-half the scaled break area associated with a 200 percent cold leg break test. The reduction in break area significantly reduced the core flow reversal that took place immediately after rupture and resulted in periods of positive core flow in the early portion of the test. The results obtained from this test are compared with results obtained from a 200 percent cold leg break test and the effect of core flow on early core thermal response is evaluated. Since Test S-29-1 was the first integral blowdown reflood test conducted with the Mod-1 system, data are also presented through the reflood stage of the test and the results are analyzed. The test data and the core thermal response calculated with the RELAP4 code are also compared

  3. A Lift-Off-Tolerant Magnetic Flux Leakage Testing Method for Drill Pipes at Wellhead.

    Science.gov (United States)

    Wu, Jianbo; Fang, Hui; Li, Long; Wang, Jie; Huang, Xiaoming; Kang, Yihua; Sun, Yanhua; Tang, Chaoqing

    2017-01-21

    To meet the great needs for MFL (magnetic flux leakage) inspection of drill pipes at wellheads, a lift-off-tolerant MFL testing method is proposed and investigated in this paper. Firstly, a Helmholtz coil magnetization method and the whole MFL testing scheme are proposed. Then, based on the magnetic field focusing effect of ferrite cores, a lift-off-tolerant MFL sensor is developed and tested. It shows high sensitivity at a lift-off distance of 5.0 mm. Further, the follow-up high repeatability MFL probing system is designed and manufactured, which was embedded with the developed sensors. It can track the swing movement of drill pipes and allow the pipe ends to pass smoothly. Finally, the developed system is employed in a drilling field for drill pipe inspection. Test results show that the proposed method can fulfill the requirements for drill pipe inspection at wellheads, which is of great importance in drill pipe safety.

  4. A Lift-Off-Tolerant Magnetic Flux Leakage Testing Method for Drill Pipes at Wellhead

    Directory of Open Access Journals (Sweden)

    Jianbo Wu

    2017-01-01

    Full Text Available To meet the great needs for MFL (magnetic flux leakage inspection of drill pipes at wellheads, a lift-off-tolerant MFL testing method is proposed and investigated in this paper. Firstly, a Helmholtz coil magnetization method and the whole MFL testing scheme are proposed. Then, based on the magnetic field focusing effect of ferrite cores, a lift-off-tolerant MFL sensor is developed and tested. It shows high sensitivity at a lift-off distance of 5.0 mm. Further, the follow-up high repeatability MFL probing system is designed and manufactured, which was embedded with the developed sensors. It can track the swing movement of drill pipes and allow the pipe ends to pass smoothly. Finally, the developed system is employed in a drilling field for drill pipe inspection. Test results show that the proposed method can fulfill the requirements for drill pipe inspection at wellheads, which is of great importance in drill pipe safety.

  5. Development, manufacturing and testing of a gas-loaded variable conductance methanol heat pipe

    Science.gov (United States)

    Vanbuggenum, R. I. J.; Daniels, D. H. W.

    1987-02-01

    The experimental technology required to measure the performance of moderate temperature heat pipes is presented. The heat pipe manufacturing process is described. The hydrodynamic characteristics of the porous structure inside the heat pipe envelope were examined using a specially developed test rig, based upon a steady-state evaporation test. A fully automated test facility was developed and validated by testing constant conductance and variable conductance heat pipes (VCHP). Theoretical performance predictions are illustrated in terms of pressure, depicted in 3D-plots, and compared with the test results of the heat pipe performance tests. The design of the VCHP was directed towards the verification of the VCHP mathematical model. The VCHP design is validated and ready for the final testing and model verification.

  6. Leak-thight seals got high pressure testing of pipes, tanks, valves

    International Nuclear Information System (INIS)

    Estrade, J.

    1985-01-01

    Leak-tight seals ensure quick, safe and efficient testing of pipes with plain-ended or flanged openings, valves with flanged or welded edges, manifields, recipients, etc. They are inserted into the pipe end manually then simply a slight turn of the seal treated wheel commences the pressure test. Hydraulic pressure is supplied by a pump through the inlet seal and air is purged through the outlet seal which then closes. The higher the pressure, the greater the sealing strength of the seal which prevents accidental unplugging. There are different types of seals: for interior plain-ended openings, for pipes with plain-ended opening, for flanged pipes. (author)

  7. Long time durability tests of fabric inlet stratification pipes

    DEFF Research Database (Denmark)

    Andersen, Elsa; Furbo, Simon

    2008-01-01

    and that this destroys the capability of building up thermal stratification for the fabric inlet stratification pipe. The results also show that although dirt, algae etc. are deposited in the fabric pipes in the space heating tank, the capability of the fabric inlet stratifiers to build up thermal stratification...

  8. Selenide isotope generator for the Galileo Mission. Axially-grooved heat pipe: accelerated life test results

    International Nuclear Information System (INIS)

    1979-08-01

    The results through SIG/Galileo contract close-out of accelerated life testing performed from June 1978 to June 1979 on axially-grooved, copper/water heat pipes are presented. The primary objective of the test was to determine the expected lifetime of axially-grooved copper/water heat pipes. The heat pipe failure rate, due to either a leak or a build-up of non-condensible gas, was determined. The secondary objective of the test was to determine the effects of time and temperature on the thermal performance parameters relevant to long-term (> 50,000 h) operation on a space power generator. The results showed that the gas generation rate appears to be constant with time after an initial sharp rise although there are indications that it drops to approximately zero beyond approx. 2000 h. During the life test, the following pipe-hours were accumulated: 159,000 at 125 0 C, 54,000 at 165 0 C, 48,000 at 185 0 C, and 8500 at 225 0 C. Heated hours per pipe ranged from 1000 to 7500 with an average of 4720. Applying calculated acceleration factors yields the equivalent of 930,000 pipe-h at 125 0 C. Including the accelerated hours on vendor tested pipes raises this number to 1,430,000 pipe-hours at 125 0 C. It was concluded that, for a heat pipe temperature of 125 0 C and a mission time of 50,000 h, the demonstrated heat pipe reliability is between 80% (based on 159,000 actual pipe-h at 125 0 C) and 98% (based on 1,430,000 accelerated pipe-h at 125 0 C). Measurements indicate some degradation of heat transfer with time, but no detectable degradation of heat transport

  9. SHAM: High-level seismic tests of piping at the HDR

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.; Malcher, L.; Schrammel, D.; Steinhilber, H.; Costello, J.F.

    1988-01-01

    As part of the second phase of vibrational/earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, high-level simulated seismic tests (SHAM) were performed during April--May 1988 on the VKL (Versuchskreislauf) in-plant piping system with two servohydraulic actuators, each capable of generating 40 tons of force. The purpose of these experiments was to study the behavior of piping subjected to seismic excitation levels that exceed design levels manifold and may result in failure/plastification of pipe supports and pipe elements, and to establish seismic margins for piping and pipe supports. The performance of six different dynamic pipe support systems was compared in these tests and the response, operability, and fragility of dynamic supports and of a typical US gate valve were investigated. Data obtained in the tests are used to validate analysis methods. Very preliminary evaluations lead to the observation that, in general, failures of dynamic supports (in particular snubbers) occur only at load levels that substantially exceed the design capacity. Pipe strains at load levels exceeding the design level threefold are quite small, and even when exceeding the design level eightfold are quite tolerable. Hence, under seismic loading, even at extreme levels and in spite of multiple support failures, pipe failure is unlikely. 5 refs., 16 figs

  10. Cryogenic and Gas System Piping Pressure Tests (A Collection of PT Permits)

    International Nuclear Information System (INIS)

    Rucinski, Russell A.

    2002-01-01

    This engineering note is a collection of pipe pressure testing documents for various sections of piping for the D-Zero cryogenic and gas systems. High pressure piping must conform with FESHM chapter 5031.1. Piping lines with ratings greater than 150 psig have a pressure test done before the line is put into service. These tests require the use of pressure testing permits. It is my intent that all pressure piping over which my group has responsibility conforms to the chapter. This includes the liquid argon and liquid helium and liquid nitrogen cryogenic systems. It also includes the high pressure air system, and the high pressure gas piping of the WAMUS and MDT gas systems. This is not an all inclusive compilation of test documentation. Some piping tests have their own engineering note. Other piping section test permits are included in separate safety review documents. So if it isn't here, that doesn't mean that it wasn't tested. D-Zero has a back up air supply system to add reliability to air compressor systems. The system includes high pressure piping which requires a review per FESHM 5031.1. The core system consists of a pressurized tube trailer, supply piping into the building and a pressure reducing regulator tied into the air compressor system discharge piping. Air flows from the trailer if the air compressor discharge pressure drops below the regulator setting. The tube trailer is periodically pumped back up to approximately 2000 psig. A high pressure compressor housed in one of the exterior buildings is used for that purpose. The system was previously documented, tested and reviewed for Run I, except for the recent addition of piping to and from the high pressure compressor. The following documents are provided for review of the system: (1) Instrument air flow schematic, drg. 3740.000-ME-273995 rev. H; (2) Component list for air system; (3) Pressure testing permit for high pressure piping; (4) Documentation from Run I contained in D-Zero Engineering note

  11. RELAP5 analysis of PKL, main steam line break test

    Energy Technology Data Exchange (ETDEWEB)

    Jonnet, J.R.; Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; With, A. de; Wakker, P.H.

    2013-12-15

    Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

  12. High temperature superconducting current lead test facility with heat pipe intercepts

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.; Prenger, C.; Roth, E.W.; Stewart, J.A.

    1998-01-01

    A high temperature superconducting (HTS) current lead test facility using heat pipe thermal intercepts is under development at the Superconducting Technology Center at Los Alamos National Laboratory. The facility can be configured for tests at currents up to 1,000 A. Mechanical cryocoolers provide refrigeration to the leads. Electrical isolation is maintained by intercepting thermal energy from the leads through cryogenic heat pipes. HST lead warm end temperature is variable from 65 K to over 90 K by controlling heat pipe evaporator temperature. Cold end temperature is variable up to 30 K. Performance predictions in terms of heat pipe evaporator temperature as a function of lead current are presented for the initial facility configuration, which supports testing up to 200 A. Measurements are to include temperature and voltage gradient in the conventional and HTS lead sections, temperature and heat transfer rate in the heat pipes. as well as optimum and off-optimum performance of the conventional lead sections

  13. Valve for the mechanical isolation of a pipe to take up a test probe

    International Nuclear Information System (INIS)

    Uecker, D.F.

    1976-01-01

    A valve is introduced for application in a pipe in which a test probe is arranged. The valve serves to isolate the pipe in a gas-tight way, thus preventing the escape of radioactive gas or dust during operation in a nuclear reactor. (TK) [de

  14. High-level seismic tests of piping at the HDR [Heissdampfreaktor

    International Nuclear Information System (INIS)

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

    1989-01-01

    As part of the second-phase testing at the Heissdampfreaktor (HDR) Test Facility in Kahl/Main, Federal Republic of Germany (FRG), high-level seismic experiments, designated SHAM, were performed on an in-plant piping system during the period of 19 April to 27 May 1988. The objectives of the SHAM experiments were to (1) study the response of piping subjected to seismic excitation levels that exceed design levels manifold and which may result in failure/plastification of pipe supports and pipe elements; (2) provide data for the validation of linear and nonlinear pipe response analyses; (3) compare and evaluate, under identical loading conditions, the performance of various dynamic support system, ranging from very flexible to very stiff support configurations; (4) establish seismic margins for piping, dynamic pipe supports, and pipe anchorages; and (5) investigate the response, operability, and fragility of dynamic supports and of a typical US gate valve under extreme levels of seismic excitation. A brief description of the SHAM tests is provided, followed by highlights of the test results that are given primarily in the form of maximum response values. Also presented are very limited comparisons of experimental data and pretest analytical predictions. 6 refs., 8 figs

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  16. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  17. Post test calculation of the experiment 'small break loss-of- coolant test' SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    International Nuclear Information System (INIS)

    Lischke, W.; Vandreier, B.

    1997-01-01

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory

  18. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W; Vandreier, B [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1998-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  19. Device for the automatic X-ray testing of welded joints of pipes

    International Nuclear Information System (INIS)

    Ries, K.; Hannoschieck, K.; Rozic, K.M.; Basler, G.

    1979-01-01

    The notification flows of the tested pipes determined by the ultrasonic inspection are transmitted to the X-ray film automatic charger in the X-ray test room. The roll table for the pipes from the ultrasonic inspection to the X-ray test room is provided with an arrangement for weld detection and tube lathe, so that the X-ray films can be set on the corresponding spot by means of a cantilever. (RW) [de

  20. BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

    1985-03-01

    This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

  1. Nondestructive testing during the fabrication of pressure vessels with half pipe jackets

    International Nuclear Information System (INIS)

    Scherner, D.

    1985-01-01

    The most important precondition to guarantee the optimum quality of half pipe jackets is the precise fixing and observance of the manufacturing conditions. For this reason the manufacturing conditions are explained in detail. The second important point is the test for gas tightness of the half pipe jacket system. The sources of mistakes in connection with the test for gas tightness are of fundamental importance. (orig./PW) [de

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

    International Nuclear Information System (INIS)

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

    1986-06-01

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

  3. Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Thinnes, G.L.

    1984-04-01

    EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping.

  4. Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes

    International Nuclear Information System (INIS)

    Ware, A.G.; Thinnes, G.L.

    1984-04-01

    EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping

  5. Sodium heat pipe module test for the SAFE-30 reactor prototype

    International Nuclear Information System (INIS)

    Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

    2001-01-01

    Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

  6. Review of pipe-break probability assessment methods and data for applicability to the advanced neutron source project for Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Fullwood, R.R.

    1989-04-01

    The Advanced Neutron Source (ANS) (Difilippo, 1986; Gamble, 1986; West, 1986; Selby, 1987) will be the world's best facility for low energy neutron research. This performance requires the highest flux density of all non-pulsed reactors with concomitant low thermal inertial and fast response to upset conditions. One of the primary concerns is that a flow cessation of the order of a second may result in fuel damage. Such a flow stoppage could be the result of break in the primary piping. This report is a review of methods for assessing pipe break probabilities based on historical operating experience in power reactors, scaling methods, fracture mechanics and fracture growth models. The goal of this work is to develop parametric guidance for the ANS design to make the event highly unlikely. It is also to review and select methods that may be used in an interactive IBM-PC model providing fast and reasonably accurate models to aid the ANS designers in achieving the safety requirements. 80 refs., 7 figs

  7. Pipe robots for internal inspection, non-destructive testing and machining of pipelines

    International Nuclear Information System (INIS)

    Reiss, Alexander

    2016-01-01

    Inspector Systems is a specialist in manufacturing of tethered self-propelled pipe robots for internal inspection, non-destructive testing and machining of pipeline systems. Our industrial sectors, which originates from 30 year experience in the nuclear industry, are Gas and Oil (On-/Offshore, Refineries), Chemical, Petrochemical, Water etc. The pipe robots are able to get inserted through poor access points (e.g. valves) and to pass in bi-directional travelling vertical sections and numerous bends with small arc radius. The paper describes the system concept and performance of the pipe robot technology. A modular construction allows to equip the robots with different operational elements for the respective application.

  8. Special fracture mechanics specimens for multilayer plastic pipes testing

    Czech Academy of Sciences Publication Activity Database

    Hutař, Pavel; Šestáková, Lucie; Knésl, Zdeněk; Nezbedová, E.; Náhlík, Luboš

    2009-01-01

    Roč. 28, č. 8 (2009), s. 785-792 ISSN 0142-9418 R&D Projects: GA ČR GA106/09/0279; GA ČR GC101/09/J027 Institutional research plan: CEZ:AV0Z20410507 Keywords : Multilayer plastic pipes * C-type specimen * K-calibration * Fracture toughness * Slow crack growth * Non-homogenous specimens Subject RIV: JL - Material s Fatigue, Friction Mechanics Impact factor: 1.667, year: 2009

  9. 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)

  10. Radiant heating tests of several liquid metal heat-pipe sandwich panels

    International Nuclear Information System (INIS)

    Camarda, C.J.; Basiulis, A.

    1983-08-01

    Integral heat pipe sandwich panels, which synergistically combine the thermal efficiency of heat pipes and the structural efficiency of honeycomb sandwich construction, were conceived as a means of alleviating thermal stress problems in the Langley Scramjet Engine. Test panels which utilized two different wickable honeycomb cores, facesheets with screen mesh sintered to the internal surfaces, and a liquid metal working fluid (either sodium or potassium) were tested by radiant heating at various heat load levels. The heat pipe panels reduced maximum temperature differences by 31 percent with sodium working fluid and 45 percent with potassium working fluid. Results indicate that a heat pipe sandwich panel is a potential, simple solution to the engine thermal stress problem. Other interesting applications of the concept include: cold plates for electronic component and circuit card cooling, radiators for large space platforms, low distortion large area structures (e.g., space antennas) and laser mirrors

  11. Experimental electro-thermal method for nondestructively testing welds in stainless steel pipes

    International Nuclear Information System (INIS)

    Green, D.R.

    1979-01-01

    Welds in austenitic stainless steel pipes are notoriously difficult to nondestructively examine using conventional ultrasonic and eddy current methods. Survace irregularities and microscopic variations in magnetic permeability cause false eddy current signal variations. Ultrasonic methods have been developed which use computer processing of the data to overcome some of the problems. Electro-thermal nondestructive testing shows promise for detecting flaws that are difficult to detect using other NDT methods. Results of a project completed to develop and demonstrate the potential of an electro-thermal method for nondestructively testing stainless steel pipe welds are presented. Electro-thermal NDT uses a brief pulse of electrical current injected into the pipe. Defects at any depth within the weld cause small differences in surface electrical current distribution. These cause short-lived transient temperature differences on the pipe's surface that are mapped using an infrared scanning camera. Localized microstructural differences and normal surface roughness in the welds have little effect on the surface temperatures

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, Thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCE is expected to closely model a LPWR LOCA. 2 - Description of test: Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development. The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs). This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR. Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coast down, and United Kingdom minimum safeguard emergency core coolant injection rates. The nuclear fuel rods were not pressurized. The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  14. 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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Key quality aspects for a new metallic composite pipe: corrosion testing, welding, weld inspection and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Conder, Robert J.; Felton, Peter [Xodus Group Ltd., Aberdeen (United Kingdom); Smith, Richard [Shell Global Solutions Inc., Houston, TX (United States); Burke, Raymond [Pipestream Inc., Houston, TX (United States); Dikstra, Frits; Deleye, Xavier [Applus RTD Ltd., Rotterdam (Netherlands)

    2010-07-01

    XPipeTM is a new metallic composite pipe. This paper discusses three aspects of this new technology. The first subject is determination of the probability of hydrogen embrittlement by the XPipeTM manufacturing method. Two materials were analyzed in three tests: slow strain rate test, constant load test and notched tensile test. The results showed that the high strength steels used do not appear to be susceptible to hydrogen embrittlement. The second subject of this article is weld inspection. A non-destructive testing method of girth welds is developed to allow inspection of the thin-walled austenitic liner pipe. The results demonstrated that the welds can be inspected using the creeping wave technique. The third subject is quality control systems using the SCADA system, which maintains traceability of the materials and monitors and records all parameters during the production process. This system appears to be efficient in ensuring that the product pipe meets recognized quality standards.

  17. Assessment of grass root effects on soil piping in sandy soils using the pinhole test

    Science.gov (United States)

    Bernatek-Jakiel, Anita; Vannoppen, Wouter; Poesen, Jean

    2017-10-01

    Soil piping is an important land degradation process that occurs in a wide range of environments. Despite an increasing number of studies on this type of subsurface erosion, the impact of vegetation on piping erosion is still unclear. It can be hypothesized that vegetation, and in particular plant roots, may reduce piping susceptibility of soils because roots of vegetation also control concentrated flow erosion rates or shallow mass movements. Therefore, this paper aims to assess the impact of grass roots on piping erosion susceptibility of a sandy soil. The pinhole test was used as it provides quantitative data on pipeflow discharge, sediment concentration and sediment discharge. Tests were conducted at different hydraulic heads (i.e., 50 mm, 180 mm, 380 mm and 1020 mm). Results showed that the hydraulic head was positively correlated with pipeflow discharge, sediment concentration and sediment discharge, while the presence of grass roots (expressed as root density) was negatively correlated with these pipeflow characteristics. Smaller sediment concentrations and sediment discharges were observed in root-permeated samples compared to root-free samples. When root density exceeds 0.5 kg m- 3, piping erosion rates decreased by 50% compared to root-free soil samples. Moreover, if grass roots are present, the positive correlation between hydraulic head and both sediment discharge and sediment concentration is less pronounced, demonstrating that grass roots become more effective in reducing piping erosion rates at larger hydraulic heads. Overall, this study demonstrates that grass roots are quite efficient in reducing piping erosion rates in sandy soils, even at high hydraulic head (> 1 m). As such, grass roots may therefore be used to efficiently control piping erosion rates in topsoils.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  19. 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

  20. 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

  1. 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.

  2. SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

    2012-10-15

    APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

  3. Inspection of piping wall loss with flow accelerated corrosion accelerated simulation test

    International Nuclear Information System (INIS)

    Ryu, Kyung Ha; Kim, Ji Hak; Hwang, Il Soon; Lee, Na Young; Kim, Ji Hyun

    2009-01-01

    Flow Accelerated Corrosion (FAC) has become a hot issue for aging of passive components. Ultrasonic Technique (UT) has been adopted to inspect the secondary piping of Nuclear Power Plants (NPPs). UT, however, uses point detection method, which results in numerous detecting points and thus takes time. We developed an Equipotential Switching Direct Current Potential Drop (ES-DCPD) method to monitor the thickness of piping that covers wide range of piping at once time. Since the ES-DCPD method covers area, not a point, it needs less monitoring time. This can be a good approach to broad carbon steel piping system such as secondary piping of NPPs. In this paper, FAC accelerated simulation test results is described. We realized accelerated FAC phenomenon by 2 times test: 23.7% thinning in 216.7 hours and 51% thinning in 795 hours. These were monitored by ES-DCPD and traditional UT. Some parameters of water chemistry are monitored and controlled to accelerate FAC process. As sensitive factors on FAC, temperature and pH was changed during the test. The wall loss monitored results reflected these changes of water chemistry successfully. Developed electrodes are also applied to simulation loop to monitor water chemistry. (author)

  4. Simulation and analysis of a main steam line transient with isolation valves closure and subsequent pipe break

    Energy Technology Data Exchange (ETDEWEB)

    Stevanovic, Vladimir; Studovic, Milovan [Faculty of Mechanical Engineering, University of Belgrade, Belgrade (Yugoslavia); Bratic, Aleksandar [Thermal Power Plant Nikola Tesla (Yugoslavia)

    1993-11-01

    Simulation and analysis of a real main steam line break transient at the coal fired 300 MW Drmno Thermal Power Plant have been performed by the computer code TEA-01. The methods and procedures used could be applied to a nuclear power plant. 9 refs., 6 figs.

  5. 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

  6. 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.)

  7. Accelerated life tests of specimen heat pipe from Communication Technology Satellite (CTS) project

    Science.gov (United States)

    Tower, L. K.; Kaufman, W. B.

    1977-01-01

    A gas-loaded variable conductance heat pipe of stainless steel with methanol working fluid identical to one now on the CTS satellite was life tested in the laboratory at accelerated conditions for 14 200 hours, equivalent to about 70 000 hours at flight conditions. The noncondensible gas inventory increased about 20 percent over the original charge. The observed gas increase is estimated to increase operating temperature by about 2.2 C, insufficient to harm the electronic gear cooled by the heat pipes in the satellite. Tests of maximum heat input against evaporator elevation agree well with the manufacturer's predictions.

  8. 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.

  9. Scientific design of the test facility for the KNGR DVI line small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

    1999-03-01

    Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

  10. Stress corrosion evaluation of powder metallurgy aluminum alloy 7091 with the breaking load test method

    Science.gov (United States)

    Domack, Marcia S.

    1987-01-01

    The stress corrosion behavior of the P/M aluminum alloy 7091 is evaluated in two overaged heat treatment conditions, T7E69 and T7E70, using an accelerated test technique known as the breaking load test method. The breaking load data obtained in this study indicate that P/M 7091 alloy is highly resistant to stress corrosion in both longitudinal and transverse orientations at stress levels up to 90 percent of the material yield strength. The reduction in mean breaking stress as a result of corrosive attack is smallest for the more overaged T7E70 condition. Details of the test procedure are included.

  11. 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

  12. 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

  13. Design, fabrication and performance tests for a polymer-based flexible flat heat pipe

    International Nuclear Information System (INIS)

    Hsieh, Shou-Shing; Yang, Ya-Ru

    2013-01-01

    Highlights: ► Fabrication of a polymer-based flexible flat heat pipe. ► Bending angle of 15° will lead to a better thermal performance of the system. ► Powers higher than 12.67 W can be transferred/delivered. - Abstract: In this paper, we report on the novel design, fabrication and performance tests for a polymer-based flexible flat heat pipe (FHP) with a bending angle in the range of 15–90°. Each heat pipe is 4 mm thick, 20 mm wide and 80 mm long, with two layers of No. 250 copper mesh as the wicking material. A copper/silicone rubber hybrid structure is designed and fabricated to achieve the flexibility of the heat pipe. Thermal characteristics are measured and studied for de-ionized water under different working conditions. Experimental results reveal that a bending angle of 15° on the vertical plane has a better thermal performance than those of heat pipes with/without bending. In addition, a higher power of 12.67 W can be transferred/delivered

  14. On the impact bending test technique for high-strength pipe steels

    Science.gov (United States)

    Arsenkin, A. M.; Odesskii, P. D.; Shabalov, I. P.; Likhachev, M. V.

    2015-10-01

    It is shown that the impact toughness (KCV-40 = 250 J/cm2) accepted for pipe steels of strength class K65 (σy ≥ 550 MPa) intended for large-diameter gas line pipes is ineffective to classify steels in fracture strength. The results obtained upon testing of specimens with a fatigue crack and additional sharp lateral grooves seem to be more effective. In energy consumption, a macrorelief with splits is found to be intermediate between ductile fracture and crystalline brittle fracture. A split formation mechanism is considered and a scheme is proposed for split formation.

  15. Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

    Energy Technology Data Exchange (ETDEWEB)

    Tarallo, Andrea, E-mail: andrea.tarallo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Mozzillo, Rocco; Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Aiello, Antonio; Utili, Marco [ENEA UTIS, C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ricapito, Italo [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • The use of human modeling tools for piping design in view of maintenance is discussed. • A possible preliminary layout for TBM subsystems in CVCS area has been designed with CATIA. • A DHM-based method to quickly check for maintainability of piping systems is suggested. - Abstract: This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

  16. ANALYSIS OF MATERIALS IN AN EXPERIMENTAL TESTING PIPE SYSTEM FOR AN INHIBITOR OF MUSSEL KILL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel P. Molloy

    2003-06-04

    A comprehensive series of 16 laboratory experiments demonstrated that the presence of vinyl tubing within a recirculating pipe system was responsible for lowering zebra mussel kill following treatment with the bacterium Pseudomonas fluorescens. All vinyl tubing was replaced in all testing units with silicone tubing, and high mussel kill (>95%) was then obtained.

  17. Testing Methodology of Breaking into Secured Storages of Mobile Operational System Google Android

    Directory of Open Access Journals (Sweden)

    Elena Vyacheslavovna Elistratova

    2013-02-01

    Full Text Available The methodology is developed for carrying out the test of breaking into internal storages of mobile operational system Google Android in order to detect security threats for personal data.

  18. Restart Testing Program for piping following steam generator replacement at North Anna Unit 1

    International Nuclear Information System (INIS)

    Bain, R.A.; Bayer, R.K.

    1993-01-01

    In order to provide assurance that the effects of performing steam generator replacement (SGR) at North Anna unit 1 had no adverse impact on plant piping systems, a cold functional verification restart testing program was developed. This restart testing program was implemented in lieu of a hot functional testing program normally used during the initial startup of a nuclear plant. A review of North Anna plant-specific and generic U.S. Nuclear Regulatory Commission requirements for restart testing was performed to ensure that no mandatory hot functional testing was required. This was determined to be the case, and the development of a cold functional test program was initiated. The cold functional test had inherent advantages as compared to the hot functional testing, while still providing assurance of piping system adequacy. The advantages of the cold verification program included reducing risk to personnel from hot piping, increasing the accuracy of measurements with the improvement in work conditions, eliminating engineering activities during the heatup process, and being able to record measurements as construction work was completed allowing for rework or repair of components if required. To ensure the effectiveness of the cold verification program, a project procedure was generated to identify the personnel, equipment, and measurement requirements. An engineering calculation was issued to document the scope of the restart test program, and an additional calculation was developed to provide acceptance criteria for the critical commodity measurements

  19. Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

    Science.gov (United States)

    Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

    2013-01-01

    Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

  20. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  1. Accelerated corrosion test for metal drainage pipes : final report.

    Science.gov (United States)

    1987-06-01

    This study represents an attempt to develop an accelerated test which would assist the highway engineer in evaluating the usefulness of a new type of coated steel culvert. The test method was to be short in duration (in the order of days), and the re...

  2. Correlation of analysis with high level vibration test results for primary coolant piping

    International Nuclear Information System (INIS)

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

    1992-01-01

    Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  4. Further development and test of the calculation code PROST for the probabilistic and deterministic evaluation of piping and vessels in the frame of leak-before-break confirmation; Weiterentwicklung und Erprobung des Rechenprogramms PROST zur probabilistischen und deterministischen Bewertung von Rohrleitungen und Behaeltern im Rahmen von Leck-vor-Bruch Nachweisen

    Energy Technology Data Exchange (ETDEWEB)

    Heckmann, Klaus; Sievers, Juergen; Arndt, Jens; Blaesius, Christoph

    2016-07-15

    The structural mechanics computer code PROST of GRS was extended in the project RS1516 for more applications and improved usability. The fracture mechanical features were extended by supplementary structure types, additional crack geometries, refined failure criteria, and temperature-dependent parameters. The consideration of in-service inspections was revised, while upgraded fatigue analyses enhance the quality of the computed results. The modeling of leakage rates was ameliorated by coupling the WinLeck computer code of GRS for leak rate computation, leading to better predictions of these features. Additional probabilistic techniques were implemented in PROST, which allow a more accurate quantification of risks. PROST's validation was done within in the project by two international benchmark studies. Analytical techniques for fracture mechanical parameters and in-service inspections in probabilistic crack-growth analyses were in the focus of the comparative studies. In addition, diverse test cases are analyzed, allowing the comparison with published studies. The available methodology for leak rate analysis was used for the planning of upcoming measurements. The PROST code has a graphical user interface, is accompanied by a detailed documentation and is made available for external users. Thus, it is a qualified tool for fracture mechanical studies, taking into account nuclear safety standards.

  5. Effects of Cross-Linking on the Hydrostatic Pressure Testing for HDPE Pipe Material using Electron Beam Machine

    International Nuclear Information System (INIS)

    Mohd Jamil Bin Hashim

    2011-01-01

    One of the most inventive, sustainable strategies used in engineering field is to improve the quality of material and minimize production cost of material for example in this paper is HDPE material. This is because HDPE is an oil base material. This paper proposes to improve its hydrostatic pressure performance for HDPE pipe. The burst test is the most direct measurement of a pipe materials resistance to hydrostatic pressure. Test will be conducted in accordance with ASTM standard for HDPE pipe that undergo electron beam irradiation cross-linking. Studies show the effect of electron beam irradiation will improve the mechanical properties of HDPE pipe. When cross-linking is induced, the mechanical properties such as tensile strength and young modulus is increase correspond to the radiation dose. This happen because the structure of HDPE, which is thermoplastic change to thermosetting. This will indicate the variability of irradiation dose which regard to the pipe pressure rating. Hence, the thickness ratio of pipe will be re-examining in order to make the production of HDPE pipe become more economical. This research review the effects of electron beam on HDPE pipe, as well as to reduce the cost of its production to improve key properties of selected plastic pipe products. (author)

  6. Phased array ultrasonic testing of dissimilar metal pipe weld joints

    International Nuclear Information System (INIS)

    Rajeev, J.; Sankaranarayanan, R.; Sharma, Govind K; Joseph, A.; Purnachandra Rao, B.

    2015-01-01

    Dissimilar metal weld (DMW) joints made of stainless steel and ferritic steel is used in nuclear industries as well as oil and gas industries. These joints are prone to frequent failures which makes the non-destructive testing of dissimilar metal weld joints utmost important for reliable and safe operation of nuclear power plants and oil and gas industries. Ultrasonic inspection of dissimilar metal weld joints is still challenging due to the inherent anisotropic and highly scattering nature. Phased array ultrasonic testing (PAUT) is an advanced technique and its capability has not been fully explored for the inspection of dissimilar metal welds

  7. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

  8. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

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

    1994-10-01

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

  9. Testing in support of on-site storage of residues in the Pipe Overpack Container

    International Nuclear Information System (INIS)

    Ammerman, D.J.; Bobbe, J.G.; Arviso, M.

    1997-02-01

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plans call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. It is desirable to use this same waste packaging for interim on-site storage in non-hardened buildings. To meet the safety concerns for this storage the Pipe Overpack Container has been subjected to a series of tests at Sandia National Laboratories in Albuquerque, New Mexico. In addition to the tests required to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II several tests were performed solely for the purpose of qualifying the container for interim storage. This report will describe these tests and the packages response to the tests. 12 figs., 3 tabs

  10. A Nonlinear Unit Root Test in the Presence of an Unknown Break

    OpenAIRE

    Popp, Stephan

    2008-01-01

    The Perron test is the most commonly applied procedure to test for a unit root in the presence of a structural break of unknown timing in the trend function. Deriving the Perron-type test regression from an unobserved component model, it is shown that the test regression in fact is nonlinear in coefficient. Taking account of the nonlinearity leads to a test with properties that are exclusively assigned to Schmidt-Phillips LM-type unit root tests.

  11. 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

  12. Comparison of fracture toughness values from large-scale pipe system tests and C(T) specimens

    International Nuclear Information System (INIS)

    Olson, R.; Scott, P.; Marschall, C.; Wilkowski, G.

    1993-01-01

    Within the International Piping Integrity Research Group (IPIRG) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe and the experiments were conducted at 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment of standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria. 4 refs., 14 figs., 1 tab

  13. Study of the performances of acoustic emission testing for glass fibre reinforced plastic pipes containing defects

    International Nuclear Information System (INIS)

    Villard, D.; Vidal, M.C.

    1995-08-01

    Glass fibre reinforced plastic pipes are more and more often used, in nuclear power plants, for building or replacement of water pipings classified 'nuclear safety'. Tests have been performed to evaluate the performances of acoustic emission testing for in service inspection of these components. The tests were focused on glass fibre reinforced polyester and vinyl-ester pipes, in as received conditions or containing impacts, and intentionally introduced defects. They have been carried out by CETIM, following the ASTM Standard E 1118 (code CARP), to a maximum pressure lever of 25 Bar The results show that the CARP procedure can be used for detection of defects and evaluation of their noxiousness towards internal pressure: most of the tubes containing low energy impacts could not be distinguished from tubes without defect; on the other hand the important noxiousness of lacks of impregnation of roving layer appeared clearly. Complementary tests have been performed on some tubes at a more important pressure lever, for which the damage of the tubes in enough to deteriorate there elastic properties. The results showed that CARP procedure give valuable informations on damage level. It would be interesting to evaluate acoustic emission on tubes containing realistic in-service degradations. (author). 11 refs., 15 figs., 6 tabs., 2 appends

  14. Laboratory piping system vibration tests to determine parametric effects on damping in the seismic frequency range

    International Nuclear Information System (INIS)

    Ware, A.G.

    1987-01-01

    A pipe damping research program is being conducted for the United States Nuclear Regulatory Commission at the Idaho National Engineering Laboratory to establish more realistic, best-estimate damping values for use in dynamic structural analyses of piping systems. As part of this program, tests were conducted on a 5-in. (128 mm ID) laboratory piping system to determine the effects of pressure, support configuration, insulation and response amplitude on damping. The tests were designed to produce a wide range of damping values, from very low damping in lightly excited uninsulated systems with few supports, to higher damping under conditions of either/or insulation, high level excitation, and various support arrangements. The effect of pressure at representative seismic levels was considered to be minimal. The supports influence damping at all excitation levels; damping was highest when a mechanical snubber was present in the system. The addition of insulation produced a large increase in damping for the hydraulic shaker excitation tests, but there was no comparable increase for the snapback excitation tests. Once a response amplitude of approximately one-half yield stress was reached, overall damping increased to relatively high levels (>10% of critical)

  15. 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)

  16. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  17. 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

  18. Results of Caisson Breakwater Tests in Multidirectional Breaking Seas

    DEFF Research Database (Denmark)

    Kofoed, Jens Peter

    In extension of the work done at the Hydraulics and Coastal Engineering Laboratory, Dept. of Civil Engineering, Aalborg University, Denmark, described in Grønbech et al., 1997, laboratory tests has been performed investigating wave loading and overtopping on caisson breakwaters in multidirectional...

  19. Calculations of Edwards' pipe blowdown tests using the code TRAC P1

    International Nuclear Information System (INIS)

    O'Mahoney, R.

    1979-05-01

    The paper describes the results obtained using the non-thermal equilibrium LOCA code TRAC-P1 for two of a series of Pipe Blowdown Tests. Comparisons are made with the experimental values and RELAP-UK Mark IV predictions. Some discrepancies between prediction and experiment are observed, and certain aspects of the model are considered to warrant possible further attention. (U.K.)

  20. Break the fast? Update on patient preparation for cholesterol testing.

    Science.gov (United States)

    Naugler, Christopher; Sidhu, Davinder

    2014-10-01

    To provide an update on the clinical usefulness of nonfasting versus fasting lipid testing to improve patient compliance, patient safety, and clinical assessment in cholesterol testing. Recommendations are identified as supported by good, fair, and poor (conflicting or insufficient) evidence, according to the classifications adopted by the Canadian Task Force on Preventive Health Care. Screening for dyslipidemia as a risk factor for coronary artery disease and management of lipid-lowering medications are key parts of primary care. Recent evidence has questioned the fasting requirement for lipid testing. In population-based studies, total cholesterol, high-density lipoprotein cholesterol, and non-low-density lipoprotein cholesterol all varied by an average of 2% with fasting status. For routine screening, nonfasting cholesterol measurement is now a reasonable alternative to a fasting cholesterol measurement. For patients with diabetes, the fasting requirement might be an important safety issue because of problems with hypoglycemia. For the monitoring of triglyceride and low-density lipoprotein cholesterol levels in patients taking lipid-lowering medications, fasting becomes more important. Fasting for routine lipid level determinations is largely unnecessary and unlikely to affect patient clinical risk stratification, while nonfasting measurement might improve patient compliance and safety. Copyright© the College of Family Physicians of Canada.

  1. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

  2. Comparison of a nonlinear dynamic model of a piping system to test data

    International Nuclear Information System (INIS)

    Blakely, K.D.; Howard, G.E.; Walton, W.B.; Johnson, B.A.; Chitty, D.E.

    1983-01-01

    Response of a nonlinear finite element model of the Heissdampfreaktor recirculation piping loop (URL) was compared to measured data, representing the physical benchmarking of a nonlinear model. Analysis-test comparisons of piping response are presented for snapback tests that induced extreme nonlinear behavior of the URL system. Nonlinearities in the system are due to twelve swaybraces (pipe supports) that possessed nonlinear force-deflection characteristics. These nonlinearities distorted system damping estimates made by using the half-power bandwidth method on Fourier transforms of measured accelerations, with the severity of distortion increasing with increasing degree of nonlinearity. Time domain methods, which are not so severely affected by the presence of nonlinearities, were used to compute system damping ratios. Nonlinear dynamic analyses were accurately and efficiently performed using the pseudo-force technique and the finite element program MSC/NASTRAN. Measured damping was incorporated into the model for snapback simulations. Acceleration time histories, acceleration Fourier transforms, and swaybrace force time histories of the nonlinear model, plus several linear models, were compared to test measurements. The nonlinear model predicted three-fourths of the measured peak accelerations to within 50%, half of the accelerations to within 25%, and one-fifth of the accelerations to within 10%. This nonlinear model predicted accelerations (in the time and frequency domains) and swaybrace forces much better than did any of the linear models, demonstrating the increased accuracy resulting from properly simulating nonlinear support behavior. In addition, earthquake response comparisons were made between the experimentally validated nonlinear model and a linear model. Significantly lower element stresses were predicted for the nonlinear model, indicating the potential usefulness of nonlinear simulations in piping design assessments. (orig.)

  3. Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

    1980-01-01

    A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

  4. Pile load test on large diameter steel pipe piles in Timan-Pechora, Russia

    Energy Technology Data Exchange (ETDEWEB)

    McKeown, S. [Golder Associates Inc., Houston, TX (United States); Tart, B. [Golder Associates Inc., Anchorage, AK (United States); Swartz, R. [Paragon Engineering Services Inc., Houston, TX (United States)

    1994-12-31

    Pile load testing conducted in May and June of 1993 at the Polar Lights Ardalin project in Arkangelsk province, Russia, was documented. Pile load testing was conducted to determine the ultimate and allowable pile loads for varying pile lengths and ground temperature conditions and to provide creep test data for deformation under constant load. The piles consisted of 20 inch diameter steel pipe piles driven open ended through prebored holes into the permafrost soils. Ultimate pile capacities, adfreeze bond, and creep properties observed were discussed. 10 figs., 4 tabs.

  5. A Lift-Off-Tolerant Magnetic Flux Leakage Testing Method for Drill Pipes at Wellhead

    OpenAIRE

    Wu, Jianbo; Fang, Hui; Li, Long; Wang, Jie; Huang, Xiaoming; Kang, Yihua; Sun, Yanhua; Tang, Chaoqing

    2017-01-01

    To meet the great needs for MFL (magnetic flux leakage) inspection of drill pipes at wellheads, a lift-off-tolerant MFL testing method is proposed and investigated in this paper. Firstly, a Helmholtz coil magnetization method and the whole MFL testing scheme are proposed. Then, based on the magnetic field focusing effect of ferrite cores, a lift-off-tolerant MFL sensor is developed and tested. It shows high sensitivity at a lift-off distance of 5.0 mm. Further, the follow-up high repeatabilit...

  6. Development and test of a space-reactor-core heat pipe

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Runyan, J.E.; Martinez, H.E.; Keddy, E.S.

    1983-01-01

    A heat pipe designed to meet the heat transfer requirements of a 100-kW/sub e/ space nuclear power system has been developed and tested. General design requirements for the device included an operating temperature of 1500 0 K with an evaporator radial flux density of 100 w/cm 2 . The total heat-pipe length of 2 m comprised an evaporator length of 0.3 m, a 1.2-m adiabatic section, and a condenser length of 0.5 m. A four-artery design employing screen arteries and distribution wicks was used with lithium serving as the working fluid. Molybdenum alloys were used for the screen materials and tube shell. Hafnium and zirconium gettering materials were used in connection with a pre-purified distilled lithium charge to ensure internal chemical compatibility. After initial performance verification, the 14.1-mm i.d. heat pipe was operated at 15 kW throughput at 1500 0 K for 100 hours. No performance degradation was observed during the test

  7. Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

    International Nuclear Information System (INIS)

    Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

    2005-01-01

    Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  9. Post-test analysis of semiscale large-break test S-06-3 using TRAC-PF1

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1982-01-01

    The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because increased application of the code to long transients such as small-break LOCAs required a faster-running code version. Although developed for long transients, its performance on large-break transients is still important. This paper assesses the ability of TRAC-PF1 to predict large-break-LOCA Test S-06-3 conducted in the Semiscale Mod-1 facility

  10. Development and testing of passive autocatalytic recombiners cooled by heat pipes

    International Nuclear Information System (INIS)

    Granzow, Christoph

    2012-01-01

    A severe accident in a nuclear power plant (NPP) can lead to core damage in conjunction with the release of large amounts of hydrogen. As hydrogen mitigation measure, passive autocatalytic recombiners (PARs) are used in today's pressurized water reactors. PARs recombine hydrogen and oxygen contained in the air to steam. The heat from this exothermic reaction causes the catalyst and its surroundings to heat up. If parts of the PAR heat up above the ignition temperature of the gas mixture, a spontaneous deflagration or detonation can occur. The aim of this work is the prevention of such high temperatures by means of passive cooling of the catalyst with heat pipes. Heat pipes are completely passive heat exchanger with a very high effective thermal conductivity. For a deeper understanding of the reaction kinetics at lower temperatures, single catalytic coated heat pipes are studied in a flow reactor. The development of a modular small-scale PAR model is then based on a test series with cooled catalyst sheets. Finally, the PAR model is tested inside a pressure vessel under boundary conditions similar to a real NPP. The experiments show, that the temperatures of the cooled catalytic sheets stay significantly below the temperature of the uncooled sheets and below the ignition temperature of the gas mixture under any set boundary conditions, although no significant reduction of the conversion efficiency can be observed. As a last point, a mathematical model of the reaction kinetics of the recombination process as well as a model of the fluid dynamic and thermohydraulic processes in a heat pipe are developed with the data obtained from the experiments.

  11. 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…

  12. Performance Test of the Microwave Ion Source with the Multi-layer DC Break

    International Nuclear Information System (INIS)

    Kim, Dae Il; Kwon, Hyeok Jung; Kim, Han Sung; Seol, Kyung Tae; Cho, Yong Sub

    2012-01-01

    A microwave proton source has been developed as a proton injector for the 100-MeV proton linac of the PEFP (Proton Engineering Frontier Project). On microwave ion source, the high voltage for the beam extraction is applied to the plasma chamber, also to the microwave components such as a 2.45GHz magnetron, a 3-stub tuner, waveguides. If microwave components can be installed on ground side, the microwave ion source can be operated and maintained easily. For the purpose, the multi-layer DC break has been developed. A multi-layer insulation has the arrangement of conductors and insulators as shown in the Fig. 1. For the purpose of stable operation as the multi-layer DC break, we checked the radiation of the insulator depending on materials and high voltage test of a fabricated multi-layer insulation. In this report, the details of performance test of the multi-layer DC break will be presented

  13. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  14. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  15. Numerical ductile tearing simulation of circumferential cracked pipe tests under dynamic loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Hyun Suk; Kim, Ji Soo; Ryu, Ho Wan; Kim, Yun Jae [Dept. of Mechanical Engineering, Korea University, Seoul (Korea, Republic of); Kim, Jin Weon [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    This paper presents a numerical method to simulate ductile tearing in cracked components under high strain rates using finite element damage analysis. The strain rate dependence on tensile properties and multiaxial fracture strain is characterized by the model developed by Johnson and Cook. The damage model is then defined based on the ductility exhaustion concept using the strain rate dependent multiaxial fracture strain concept. The proposed model is applied to simulate previously published three cracked pipe bending test results under two different test speed conditions. Simulated results show overall good agreement with experimental results.

  16. Piping damping tests evaluating influence of types of support and excitation

    International Nuclear Information System (INIS)

    Arendts, J.G.; Ware, A.G.; Gorman, V.W.

    1985-01-01

    The United States Nuclear Regulatory Commission and the Electric Power Research Institute have jointly sponsored construction of two laboratory piping systems at the ANCO Engineers facility in California. EG and G Idaho used the second of these systems to obtain piping system damping data using different supports and methods of excitation. The 6-in. carbon steel piping system was approximately 50 ft in length with two 3-in. branch lines. It was supported at five locations and excited using a single electrohydraulic shaker. Both random and swept sine methods of excitations were used. A variable support attached near the shaker location allowed four different configurations to be tested: a rigid strut, a mechanical snubber, a hydraulic snubber, and a rigid strut with a gap. Data were recorded for the lowest nine significant modes. Damping for the first three modes ranged for 1 to 3% of critical damping and decreased as frequency increased. The random excitation produced a slightly higher average overall damping of the system

  17. Standard practice for ultrasonic testing of the Weld Zone of welded pipe and tubing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice describes general ultrasonic testing procedures for the detection of discontinuities in the weld and adjacent heat affected zones of welded pipe and tubing by scanning with relative motion between the search unit and pipe or tube. When contact or unfocused immersion search units are employed, this practice is intended for tubular products having specified outside diameters ≥2 in. (≥50 mm) and specified wall thicknesses of 1/8to 11/16 in. (3 to 27 mm). When properly focused immersion search units are employed, this practice may also be applied to material of smaller diameter and thinner wall. Note 1—When contact or unfocused immersion search units are used, precautions should be exercised when examining pipes or tubes near the lower specified limits. Certain combinations of search unit size, frequency, thin–wall thicknesses, and small diameters could cause generation of unwanted sound waves that may produce erroneous examination results. 1.2 All surfaces of material to be examined in ...

  18. 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

  19. 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

  20. THE MAKE BREAK TEST AS A DIAGNOSTIC-TOOL IN FUNCTIONAL WEAKNESS

    NARCIS (Netherlands)

    VANDERPLOEG, RJO; OOSTERHUIS, HJGH

    Strength was measured in four major muscle groups with a hand-held dynamometer. The "make" and "break" technique was used with and without encouragement, and fatiguability was tested in patients with organic weakness and patients with functional weakness. Patients with functional weakness could be

  1. Interpretation, with respect to ASME code Case N-318, of limit moment and fatigue tests of lugs welded to pipe

    International Nuclear Information System (INIS)

    Foster, D.C.; Van Duyne, D.A.; Budlong, L.A.; Muffett, J.W.; Wais, E.A.; Streck, G.; Rodabaugh, E.C.

    1990-01-01

    Two nonmandatory ASME code cases have been used often in the evaluation of lugs on nuclear-power- plant piping systems. ASME Code Case N-318 provides guidance for evaluation of the design of rectangular cross-section attachments on Class 2 or 3 piping, and ASME Code Case N-122 provides guidance for evaluation of lugs on Class 1 piping. These code cases have been reviewed and evaluated based on available test data. The results indicate that the Code cases are overly conservative. Recommendations for revisions to the cases are presented which, if adopted, will reduce the overconservatism

  2. 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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  4. Ten Year Operating Test Results and Post-Test Analysis of a 1/10 Segment Stirling Sodium Heat Pipe, Phase III

    Science.gov (United States)

    Rosenfeld, John, H; Minnerly, Kenneth, G; Dyson, Christopher, M.

    2012-01-01

    High-temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, Stirling cycle heat sources; and with the resurgence of space nuclear power both as reactor heat removal elements and as radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly, long-term materials compatibility is being evaluated through the use of high-temperature life test heat pipes. Thermacore, Inc., has carried out a sodium heat pipe 10-year life test to establish long-term operating reliability. Sodium heat pipes have demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 87,000 hr (10 yr) at nearly 700 C. These life test results have demonstrated the potential for high-temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and post-test analysis of the heat pipe and sodium working fluid are described.

  5. Computerized tomography used in non-destructive testing of welded pipes

    Energy Technology Data Exchange (ETDEWEB)

    Iovea, M; Rizescu, C; Georgescu, G; Marinescu, A; Chitescu, P; Sava, T; Neagu, M; Avram, D [Institute of Research and Design for Electrical Engineering, ICPE - Electrostatica Splaiul Unirii 313, Sect. 3, R-74204 Bucharest (Romania)

    1997-12-31

    High quality standards in operation of National Power System is ensured by the use of high performance techniques and systems for Non-Destructive Testing (NDT). In recent years a number of new developments of the non-conventional technologies in the field of NDT have been achieved. In our laboratory there have been developed two computerized technologies using {gamma}-ray computed tomography and ultrasonic imaging methods. The standard techniques for imaging from projection data is computerized tomography. The industrial computerized tomography methods consist in the measurement of thin X - or {gamma}-ray beam attenuation when passing through some selected surface of the tested object, along several directions, so that by means of an adequate mathematical algorithm, a map of linear attenuation coefficients for the scanned surface is obtained. In fact, this map gives the density of materials occurring in the surface plane. Computerized tomography equipment, in various constructive versions, are intended for the following applications: (1) NDT in those fields requiring strict control of product quality, as for instance the nuclear energy, military industry, aeronautics, transportation fields, etc., (2) research in field of materials technology, machine engineering, metallurgy, welding, etc. This paper presents the applications of Computerized Tomography in NDT, by showing the results obtained on welded pipes, as well as the facilities offered by this method. In the final part, the paper presents the concept of a mobile tomography system for industrial pipes testing. (author). 2 figs., 7 refs.

  6. Computerized tomography used in non-destructive testing of welded pipes

    International Nuclear Information System (INIS)

    Iovea, M.; Rizescu, C.; Georgescu, G.; Marinescu, A.; Chitescu, P.; Sava, T.; Neagu, M.; Avram, D.

    1996-01-01

    High quality standards in operation of National Power System is ensured by the use of high performance techniques and systems for Non-Destructive Testing (NDT). In recent years a number of new developments of the non-conventional technologies in the field of NDT have been achieved. In our laboratory there have been developed two computerized technologies using γ-ray computed tomography and ultrasonic imaging methods. The standard techniques for imaging from projection data is computerized tomography. The industrial computerized tomography methods consist in the measurement of thin X - or γ-ray beam attenuation when passing through some selected surface of the tested object, along several directions, so that by means of an adequate mathematical algorithm, a map of linear attenuation coefficients for the scanned surface is obtained. In fact, this map gives the density of materials occurring in the surface plane. Computerized tomography equipment, in various constructive versions, are intended for the following applications: 1) NDT in those fields requiring strict control of product quality, as for instance the nuclear energy, military industry, aeronautics, transportation fields, etc., 2) research in field of materials technology, machine engineering, metallurgy, welding, etc. This paper presents the applications of Computerized Tomography in NDT, by showing the results obtained on welded pipes, as well as the facilities offered by this method. In the final part, the paper presents the concept of a mobile tomography system for industrial pipes testing. (author). 2 figs., 7 refs

  7. Presentation of accessibility equipment for primary pipings, IHX, pumps and appertaining manipulator tests

    International Nuclear Information System (INIS)

    Hahn, G.; Hoeft, E.

    1980-01-01

    Accessibility and inservice procedure of SNR-300 components are described. Due to the high radiation level in the primary system it was necessary to develop special equipment to permit access to the testing components. The pertinent examination methods for surveying welding seams are acoustic (ultrasonic) and optical procedures (TV cameras, surface crack tests). This can be done by remote-controlled manipulators and special devices, which can transport the inspection system by rails to the testing position. Presently, relatively limited experience exists for such remote-controlled handling in nuclear power plants. Thus model experiments were carried out on a model pipe section at INTERATOM. The performed test shows that the concept planned to perform inservice by using remote-controlled manipulators can be realized successfully. (author)

  8. Experiment data of 200% recirculation pump discharge line break integral test run 961 with HPCS failure at ROSA-III and comparison with results of suction line break tests

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Yonomoto, Taisuke; Shiba, Masayoshi

    1984-03-01

    This report presents the experimental data of RUN 961, a 200% double-ended break test at the recirculation pump discharge line in the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV (Main Steam Isolation Valve) closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The PCT (Peak Cladding Temperature) was 894 K, which was 107 K higher than a 200% pump suction line break test (RUN 926) due to the smaller depressurization rate. The effect of break location on transient LOCA phenomena was clarified by comparing the discharge and suction break tests. The whole core was quenched 71 s after LPCI actuation and the effectiveness of ECCS has been confirmed. (author)

  9. ROSA-III 200% double-ended break integral test RUN 901

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Tasaka, Kanji; Koizumi, Yasuo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Suzuki, Mitsuhiro; Shiba, Masayoshi

    1984-02-01

    This report presents the experimental data of RUN 901, a 200% double-ended break test at the recirculation pump suction line with the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The channel inlet flows were measured by differential pressure transducers installed at the channel inlet orifices of the fuel assembly No.4. The PCT (Peak Cladding Temperature) was 780 K occured during the blowdown phase in RUN 901. The whole core was quenched after the ECCS actuation and the effectiveness of ECCS has been confirmed. (author)

  10. APPLICATION OF STEEL PIPE PILE LOADING TESTS TO DESIGN VERIFICATION OF FOUNDATION OF THE TOKYO GATE BRIDGE

    Science.gov (United States)

    Saitou, Yutaka; Kikuchi, Yoshiaki; Kusakabe, Osamu; Kiyomiya, Osamu; Yoneyama, Haruo; Kawakami, Taiji

    Steel sheet pipe pile foundations with large diameter steel pipe sheet pile were used for the foundation of the main pier of the Tokyo Gateway bridge. However, as for the large diameter steel pipe pile, the bearing mechanism including a pile tip plugging effect is still unclear due to lack of the practical examinations even though loading tests are performed on Trans-Tokyo Bay Highway. In the light of the foregoing problems, static pile loading tests both vertical and horizontal directions, a dynamic loading test, and cone penetration tests we re conducted for determining proper design parameters of the ground for the foundations. Design parameters were determined rationally based on the tests results. Rational design verification was obtained from this research.

  11. 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

  12. Stands for testing the strength of welded pipe materials under the action of a corrosive medium

    Directory of Open Access Journals (Sweden)

    M.A. Kolodyi

    2017-12-01

    Full Text Available In order to study the features of the destruction of materials of pipelines for the transportation of oil, gas, products of processing of oil, water and other substances in the laboratory of the department of development of minerals named by prof. Bakka N.T. the complex of installations is invented, for which Ukrainian patents were obtained as utility models No. 30794, No. 52493, for the study of the working capacity of the elements of the listed pipeline systems in conditions that are as close as possible to the operational under the influence of the corrosive medium. Rotary vacuum devices were used as the basic elements of the proposed installations for testing the materials of the welded tubes for durability at single tensile and under flat stress conditions. The article presents the design of research stands for testing the durability of pipe materials and welds of pipelines using samples of materials and natural pipes (shortened under the influence of static, low cyclic and dynamic loads, and analyzes the influence of aggressive media.

  13. ATHLET calculations of the pressurizer surge line break (PH-SLB test) at the PMK-2 test facility

    International Nuclear Information System (INIS)

    Krepper, E.; Schaefer, F.

    2000-01-01

    At the Hungarian integral test facility PMK-2 a pressurizer surge line break experiment (PH-SLB test) was carried out with the PHARE 4.2.6b project. The primary objective of the test was to provide experimental data for a surge line break transient at VVER-440 reactors with reduced injection from the emergency core cooling systems (ECC). At the Institute of Safety Research calculations of the experiment were performed with the thermohydraulic computer code ATHLET, which was developed by GRS (Gesellschaft fuer Anlagen- und Reaktorsicherheit) mbH. In the context of the PHARE 4.2.6b project the Institute of Safety Research has also supplied the void fraction measurement system for the PMK-2 test facility and was involved in the evaluation of the experimental results. (orig.)

  14. Comparisons of TRAC-PD2 calculations with Semiscale Mod-3 small-break tests

    International Nuclear Information System (INIS)

    Gilbert, J.S.; Sahota, M.S.; Boyack, B.E.; Booker, C.P.; Meier, J.K.

    1981-01-01

    Five experiments conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory (INEL) were calculated using the latest released version of the Transient Reactor Analysis Code (TRAC-PD2). The results were used to assess TRAC-PD2 predictions of thermal-hydraulic phenomena and the effects of pump operation on system response during slow transients. Tests S-SB-P1, S-SB-P2, and S-SB-P7 simulated equivalent 2.5% communicative cold-leg breaks for early pump-trip (pumps-off), intermediate pump-trip (pumps-on), and late pump-trip (pumps-on) operation, respectively. Tests S-SB-P3 and S-SB-P4 simulated equivalent 2.5% communicative hot-leg breaks for pumps-off and pumps-on operation, respectively. Parameters examined in the study included primary system mass distribution, mass inventory, and void fraction distribution

  15. Standard test method for determination of breaking strength of ceramic tiles by three-point loading

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This test method covers the determination of breaking strength of ceramic tiles by three-point loading. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  16. Purchasing Power Parity in Transition Countries: Panel Stationary Test with Smooth and Sharp Breaks

    Directory of Open Access Journals (Sweden)

    Mohsen Bahmani-Oskooee

    2015-05-01

    Full Text Available This study examines whether the long-run purchasing power parity (PPP holds in transition economies (Bulgaria, the Czech Republic, Hungary, Latvia, Lithuania, Poland, Romania and Russia using monthly data over the 1995–2011 period. We apply a recently introduced panel stationary test, which accounts for sharp breaks and smooth shifts. The results indicate that the PPP holds only in two countries (i.e., Lithuania and Poland.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. Analytical study for frequency effects on the EPRI/USNRC piping component tests. Part 1: Theoretical basis and model development

    International Nuclear Information System (INIS)

    Adams, T.M.; Branch, E.B.; Tagart, S.W. Jr.

    1994-01-01

    As part of the engineering effort for the Advanced Light Water Reactor the Advanced Reactor Corporation formed a Piping Technical Core Group to develop a set of improved ASME Boiler and Pressure Vessel Code, Section III design rules and approaches for ALWR plant piping and support design. The technical basis for the proposed changes to the ASME Boiler and Pressure Vessel Code developed by Technical Core Group for the design of piping relies heavily on the failure margins determined from the EPRI/USNRC piping component test program. The majority of the component tests forming the basis for the reported margins against failure were run with input frequency to natural frequency ratios (Ω/ω) in the range of 0.74 to 0.87. One concern investigated by the Technical Core Group was the effect which could exist on measured margins if the tests had been run at higher or lower frequency ratios than those in the limited frequency ratio range tested. Specifically, the concern investigated was that the proposed Technical Core Group Piping Stress Criteria will allow piping to be designed in the low frequency range (Ω/ω ≥ 2.0) for which there is little test data from the EPRI/USNRC test program. The purpose of this analytical study was to: (1) evaluate the potential for margin variation as a function of the frequency ratio (R ω = Ω/ω, where Ω is the forcing frequency and ω is the natural component frequency), (2) recommend a margin reduction factor (MRF) that could be applied to margins determined from the EPRI/USNRC test program to adjust those margins for potential margin variation with frequency ratio. Presented in this paper is the analytical approach and methodology, which are inelastic analysis, which was the basis of the study. Also, discussed is the development of the analytical model, the procedure used to benchmark the model to actual test results, and the various parameter studies conducted

  19. A creep life assessment method for boiler pipes using small punch creep test

    International Nuclear Information System (INIS)

    Izaki, Toru; Kobayashi, Toshimi; Kusumoto, Junichi; Kanaya, Akihiro

    2009-01-01

    The small punch creep (SPC) test is considered as a highly useful method for creep life assessment for high temperature plant components. SPC uses miniature-sized specimens and does not cause any serious sampling damages, and its assessment accuracy is at a high level. However, in applying the SPC test to the residual creep life assessment of the boiler in service, there are some issues to be studied. In order to apply SPC test to the residual creep life assessment of the 2.25Cr-1Mo steel boiler pipe, the relationship between uniaxial creep stress and the SPC test load has been studied. The virgin material, pre-crept, weldment and service aged samples of 2.25Cr-1Mo steel were tested. It was confirmed that the relationship between uniaxial creep stress and the SPC test load at the same rupture time can be described as a single straight line independent of test conditions and materials. Therefore a life assessment is possible by using SPC test in place of uniaxial creep tests. The creep life assessment using SPC was applied to actual thermal power plant components which are in service.

  20. Double-Ended Break Test of an 8.5 inch Direct Vessel Injection Line using the ATLAS

    International Nuclear Information System (INIS)

    Choi, Ki Yong; Kang, Kyoung Ho; Kim, Bok Deuk; Kim, Yeon Sik; Min, Kyoung Ho; Park, Choon Kyoung; Park, Hyun Sik; Baek, Won Pil; Cho, Seok; Choi, Nam Hyun

    2010-01-01

    A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at the Korea Atomic Energy Research Institute (KAERI). It is a 1/2 reduced height and 1/288-volume scaled test facility with respect to the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. In 2008, an integral effect test for simulating a guillotine break of a DVI line of the APR1400 was carried out as the first DVI test item, named as SB-DVI-03, on May, 2008. With an improvement on the break flow measuring system, the second DVI test for a guillotine break of a DVI line of the APR1400, named as SB-DVI-08, was conducted for repeatability. The present data is the first integral effect test data of its kind for simulating a DVI line break accident. It will help in understanding the thermal hydraulic phenomena occurring during the DVI line break accident. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. The present integral effect test data will be used to validate the current safety analysis methodology for the DVI line break accident

  1. Real-time numerical evaluation of dynamic tests with sudden closing of valves in piping systems

    International Nuclear Information System (INIS)

    Geidel, W.; Leimbach, K.R.

    1979-01-01

    The sudden closing of a valve in a piping system causes a build-up of pressure which, in turn, causes severe vibrations of the structural system. The licensing procedure calls for on-site tests to determine the dynamic effects of such closing of valves, and to check the stresses and displacements against the allowable ones. The measurements include time histories of displacements, accelerations and internal pressure. The computer program KWUROHR for the static and dynamic analysis of piping systems has been used by KWU and several subcontractors during the past four vears. This program has been extended by adding a subroutine package which computes time histories of displacements, accelerations and stresses resulting from the input of measured time histories of internal pressures at selected locations. The computer algorithm establishes the topological connectivity between the internal pressure measuring locations, to set up a logic for linear pressure interpolation between these points and pressure steps at reducers and valves. A minimum number of input points is required to give realistic results. (orig.)

  2. Testing in support of transportation of residues in the pipe overpack container

    International Nuclear Information System (INIS)

    Ammerman, D.J.; Bobbe, J.G.; Arviso, M.; Bronowski, D.R.

    1997-04-01

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plants call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. The tests described here were performed to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II. Using a more robust container will assure the fissile materials in each container can not be mixed with the fissile material from the other containers and will provide criticality control. This will allow an increase in the payload of the TRUPACT-II from 325 fissile gram equivalents to 2,800 fissile gram equivalents

  3. Manufacture and test of prototype water pipe chase barrier in ITER Magnet Feeder system

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Kun, E-mail: lukun@ipp.ac.cn [Institute of Plasma Physics, Shushan Hu Road 350, Hefei, Anhui (China); Wen, Xinjie; Liu, Chen; Song, Yuntao [Institute of Plasma Physics, Shushan Hu Road 350, Hefei, Anhui (China); Niu, Erwu [ITER China, 15B Fuxing Road, Beijing 100862 (China); Gung, Chenyu; Su, Man [ITER Organization, Route de Vinon-sur-Verdon – CS 90046, 13067 St Paul-lez-Durance Cedex (France)

    2016-11-01

    The Magnet Feeder system in the International Thermonuclear Experimental Reactor (ITER) deploys electrical currents and supercritical helium to the superconducting magnets and the magnet diagnostic signals to the operators. In the current design, the feeders located in the upper L3 level of the Tokamak gallery penetrate the Tokamak coolant water system vault, the biological shield and the cryostat. As a secondary confinement to contain the activated coolant water in the vault in the case of water pipe burst accident, a water barrier is welded between the penetration in the water pipe chase outer wall and the mid-plane of the vacuum jacket of the Feeder Coil Terminal Box (CTB). A thin-wall stainless steel diaphragm with an omega shape profile is welded around the CTB as the water barrier to endure 2 bar hydraulic pressure. In addition, the barrier is designed as a flexible compensator to withstand a maximum of 15 mm of axial displacement of the CTB in case of helium leak accident without failure. This paper presents the detail configuration, the manufacturing and assembly processes of the water barrier. Test results of the prototype water barrier under simulated accident conditions are also reported. Successful qualification of the design and manufacturing process of the water barrier lays a good foundation for the series production of this subsystem.

  4. Investigation into the cause of leak in the pipe of the corrosion test apparatus of IS process

    International Nuclear Information System (INIS)

    Tanaka, Nobuyuki; Furukawa, Tomohiro; Inagaki, Yoshiyuki; Suwa, Hirokazu

    2008-12-01

    The thermochemical water-splitting hydrogen production IS process utilizes corrosive chemicals such as sulfuric acid and hydriodic acid. Corrosion tests in IS process environments have been carried out to get the corrosion data of materials. In the corrosion test in 90wt% sulfuric acid at 400degC, the leak of sulfuric acid was observed in a pipe connected with a reflux condenser. The cause of the leakage is a significant knowledge for the operation of the test apparatus. Therefore the cause was investigated. A 1mm wide through hole was detected in the pipe around the welding bead. By visual observation after cutting the pipe, the wall thickness of the pipe became thin at the inside welding bead around the through hole. In addition, EMPA showed that the inhomogeneous distribution of the constituent elements of the pipe was observed around the through hole. For these reasons, it is estimated that the lowering of the corrosion resistance by the sensitization at the welding caused the leakage. (author)

  5. The Impact of Structural Break(s on the Validity of Purchasing Power Parity in Turkey: Evidence from Zivot-Andrews and Lagrange Multiplier Unit Root Tests

    Directory of Open Access Journals (Sweden)

    Hakan Kum

    2012-01-01

    Full Text Available This study examines the validity of the purchasing power parity (PPP in Turkey for annual data from 1953 to 2009. While results from both the ADF unit root and the DF-GLS unit root test indicate mixed results, PPP holds for Turkey with the presence of structural breaks which are obtained by Zivot and Andrews and Lagrange Multiplier unit root tests.

  6. Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1995-10-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ''like-new'' condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ''like-new'' condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report

  7. Design and testing of a heat pipe gas combustion system for the STM4-120 Stirling engine

    Science.gov (United States)

    Khalili, K.; Godett, T. M.; Meijer, R. J.; Verhey, R. P.

    Evaporators of a novel geometry, designed to have a more compact size yet the same output as larger, conventional heat pipes, have been fabricated and tested. A technique was developed to calculate capillary pressure required inside the heat pipe. Several quarter- and full-scale evaporators were designed and successfully tested. The burner, film-cooled combustion chamber, and preheater were designed and tested separately. A complete heat pipe gas combustion system (HPGC) was tested, showing an efficiency of 89 percent was measured at 20 kWth. A film-cooled combustion chamber was tested with flame temperatures of 2200 C and wall temperatures below 1000 C using preheated air for film cooling. Also, a full-scale HPGC was tested at an excess of 95 kWth, showing efficiency in the range of 85 to 90 percent under steady-state conditions. Results of transient and startup tests, carried out to evaluate the performance of the heat pipe, all also reported.

  8. RELAP5 simulations of critical break experiments in the RD-14 test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, I-G; Cho, Y-J; Lee, S [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1996-12-31

    RELAP5/MOD3 simulations of critical break tests in the RD-14 facility, modelling a loss of coolant in a CANDU reactor, were compared to the experimental results, and to CATHENA simulations of the early stage of the test. The RELAP5/MOD3 predicted thermal hydraulic behaviour reasonably well, but some discrepancies were observed after emergency cooling injection (ECI). Pressure differences between headers govern flow through the heated sections, particularly after ECI, and there is much uncertainty in the header pressures; further work is therefore recommended. 6 refs., 3 figs.

  9. Uplifting behavior of shallow buried pipe in liquefiable soil by dynamic centrifuge test.

    Science.gov (United States)

    Huang, Bo; Liu, Jingwen; Lin, Peng; Ling, Daosheng

    2014-01-01

    Underground pipelines are widely applied in the so-called lifeline engineerings. It shows according to seismic surveys that the damage from soil liquefaction to underground pipelines was the most serious, whose failures were mainly in the form of pipeline uplifting. In the present study, dynamic centrifuge model tests were conducted to study the uplifting behaviors of shallow-buried pipeline subjected to seismic vibration in liquefied sites. The uplifting mechanism was discussed through the responses of the pore water pressure and earth pressure around the pipeline. Additionally, the analysis of force, which the pipeline was subjected to before and during vibration, was introduced and proved to be reasonable by the comparison of the measured and the calculated results. The uplifting behavior of pipe is the combination effects of multiple forces, and is highly dependent on the excess pore pressure.

  10. Uplifting Behavior of Shallow Buried Pipe in Liquefiable Soil by Dynamic Centrifuge Test

    Directory of Open Access Journals (Sweden)

    Bo Huang

    2014-01-01

    Full Text Available Underground pipelines are widely applied in the so-called lifeline engineerings. It shows according to seismic surveys that the damage from soil liquefaction to underground pipelines was the most serious, whose failures were mainly in the form of pipeline uplifting. In the present study, dynamic centrifuge model tests were conducted to study the uplifting behaviors of shallow-buried pipeline subjected to seismic vibration in liquefied sites. The uplifting mechanism was discussed through the responses of the pore water pressure and earth pressure around the pipeline. Additionally, the analysis of force, which the pipeline was subjected to before and during vibration, was introduced and proved to be reasonable by the comparison of the measured and the calculated results. The uplifting behavior of pipe is the combination effects of multiple forces, and is highly dependent on the excess pore pressure.

  11. Uplifting Behavior of Shallow Buried Pipe in Liquefiable Soil by Dynamic Centrifuge Test

    Science.gov (United States)

    Liu, Jingwen; Ling, Daosheng

    2014-01-01

    Underground pipelines are widely applied in the so-called lifeline engineerings. It shows according to seismic surveys that the damage from soil liquefaction to underground pipelines was the most serious, whose failures were mainly in the form of pipeline uplifting. In the present study, dynamic centrifuge model tests were conducted to study the uplifting behaviors of shallow-buried pipeline subjected to seismic vibration in liquefied sites. The uplifting mechanism was discussed through the responses of the pore water pressure and earth pressure around the pipeline. Additionally, the analysis of force, which the pipeline was subjected to before and during vibration, was introduced and proved to be reasonable by the comparison of the measured and the calculated results. The uplifting behavior of pipe is the combination effects of multiple forces, and is highly dependent on the excess pore pressure. PMID:25121140

  12. Weld testing in the fabrication of large-diameter pipes; Schweissnahtpruefung bei der Fertigung von Grossrohren

    Energy Technology Data Exchange (ETDEWEB)

    Wenzel, T.; Fuchs, T.; Hassler, U.; Hanke, R. [Fraunhofer-Institut fuer Integrierte Schaltungen (IIS), Fuerth (Germany). EZRT; Matzen, H.U.; Kraemer, J. [GE Inspection Technologies, Ahrensburg (Germany); Lindenschmidt, H. [Butting, Knesebeck (Germany); Behrendt, R.; Kostka, G.; Schmitt, P. [Fraunhofer-Institut fuer Integrierte Schaltungen (IIS), Erlangen (Germany)

    2007-07-01

    Fully automatic radiographic testing of cast light metal components is a state of the art technology. The contribution describes its application in weld testing. A new method for evaluating X-rays of welds is presented which were tested using an innovative X-ray camera with maximum spatial resolution and a wide range of grey values. Further, a novel concept for handling test objects significantly shortens testing times. The pipes are not moved longitudinally; instead, the longitudinal motion is made by the X-ray emitter and sensor, which reduces the testing time by up to 30 percent. The specially developed X-ray detector has a sensitive surface of 200 mm x 50 mm with a total of 4.2 million pixels. Neither the evaluation electronics nor the light-sensitive camera chip are exposed to the direct X-radiation so that no damage will occur at photoenergies up to at least 250 keV. Many tests, e.g. according to EN 13068 and EN 462-5, have shown that the image quality in general and especially the local resolution exceeds the specifications of the EN 584 standard on weld testing with X-ray films. The pictures taken by the camera serve as input data for fully automatic evaluation. All stages of image processing implement 16-bit digitalisation depth in order to make use of the high dynamic range of gray value images. This means that in the whole processing chain, there will be no loss of information from downscaling of the gray values. In the first stage of image processing, the gray values are transformed into penetrated material thicknesses in preparation of the measurement of fault length in the direction of incidence at a later stage. In the next stage, external boundaries and the middle of the weld are detected, followed by an adaptive filtering stage. Additionally, information on the accurate location of the weld is transmitted to the control system of the mechanical parts, so that optimum positioning of the weld with respect to the camera is ensured. The adaptive filter

  13. 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.

  14. Remote controlled in-pipe manipulators for milling, welding and EC-testing, for application in BWRS

    International Nuclear Information System (INIS)

    Seeberger, E.K.

    2000-01-01

    Many pipes in power plants and industrial facilities have piping sections, which are not accessible from the outside or which are difficult to access. Accordingly, remote controlled pipe machining manipulators have been built which enable in-pipe inspection and repair. Since the 1980s, defects have been found at the Inconel welds of the RPV nozzles of boiling water reactors throughout the world. These defects comprise cracks caused by stress corrosion cracking in areas of manual welds made using the weld filler metal Inconel 182. The cracks were found in Inconel-182 buttering at the ferritic nozzles as well as in the welded joints connecting to the fully-austenitic safe ends (Inconel 600 and stainless steel). These welds are not accessible from outside. The ferritic nozzle is cladded with austenitic material on the inside. The adjacent buttering was applied manually using the weld filler metal Inconel 182. The safe end made of Inconel 600 was welded to the nozzle also using Inconel 182 as the filler metal. The repair problems for inside were solved with remote-controlled in-pipe manipulators which enable in-pipe inspection and repair. A complete systems of manipulators has been developed and qualified for application in nuclear power plants. The tasks that must be performed with this set of in-pipe manipulator are as follows: 1st step - Insertion of the milling/ET manipulator into piping to the work location; 2nd step Detection of the transition line with the ferritic measurement probe; 3rd step - Performance of a surface crack examination by eddy current (ET) method; 4th step - Milling of the groove and preparation for weld backlay and, in case of ET indications, elimination of such flaws also by milling. 5th step - Welding of backlay and/or repair weld using the GTA pulsed arc technique; 6th step - After welding it is necessary to prepare the surface for eddy current testing. A final milling inside the pipe is done with the milling manipulator to adjust the

  15. Assessment of the MARS-KS Code Using Atlas 6-inch cold leg Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kang, D. G.; Kim, J. S.; Ahn, S. H.; Seul, K. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-03-15

    An integral effect test on the SBLOCA (Small-Break Loss of Coolant Accident) aiming at 6-inch cold leg bottom break, SB-CL-09, was conducted with the Atlas on November, 13, 2009, by KAERI. In this study, the calculations using MARS-KS Vt1.2 code were conducted for 6-inch cold leg break test of Atlas (SB-CL-09) which is the second domestic standard problem (Dsp-02) to assess MARS-KS code capability to simulate the transient thermal-hydraulic behavior for SBLOCA. The steady state was determined by conducting a null transient calculation and the errors between the calculated and measured values are acceptable for almost primary/secondary system parameters. The predicted pressurizer pressure agrees relatively well with the experimental data and the predicted break flow and mass are in good agreement with experiment. In MARS-KS calculation, the decrease of core collapsed water level is predicted well in blowdown phase, but just before LSC, water level is higher than experiment. However, the sudden decrease and increase of water level is higher than experiment. However, the sudden decrease and increase of water level at the LSC are predicted qualitatively. After LSC, there is another water level dip at Sit injection time which is not in experiment. It is considered that this phenomenon is caused by rapid depressurization of downcomer due to significant condensation rate of vapor in downcomer when Sit water flows in it. For the downcomer water level is predicted well, however, it is significantly over-predicted at SIT injection time, water level is predicted well, however, it is significantly over-predicted at SIT injection time after SIT water flows in downcomer. Predicted cladding temperature generally agrees well with the experiment, while there is peak at SIT injection time in calculation which is not in experiment. The loop seals of 1A, 2B intermediate leg are cleared around 400 seconds in experiment, while only that of 1A is cleared in MARS-KS calculation at the

  16. In-plane and out-of-plane bending tests on carbon steel pipe bends

    International Nuclear Information System (INIS)

    Brouard, D.; Tremblais, A.; Vrillon, B.

    1979-01-01

    The objectives of these tests were to obtain experimental results on bends behaviour in elastic and plastic regime by in plane and out of plane bending. Results were used to improve the computer model, for large distorsion of bends, to be used in a simplified beam type computer code for piping calculations. Tests were made on type ANSI B 169 DN 5 bends in ASTM A 106 Grade B carbon steel. These tests made it possible to measure, for identical bends, in elastic regime, the flexibility factors and, in plastic regime, the total evolution in opening, in closing and out of plane. Flexibility factors of 180 0 bend without flanges are approximately the same in opening and in closing. The end effect due to flanges is not very significant, but it is important for 90 0 bends. In plastic regime, collapse loads or collapse moments of bends depends also of both the end effects and the angle bend. The end effects and the angle bend are more sensitive in opening than in closing. The interest of these tests is to procure some precise evolution curves of identical bends well characterized in geometry and metal strength, deflected in large distorsions. (orig./HP)

  17. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

  18. Whistling of pipes with narrow corrugations: scale model tests and consequences for carcass design

    NARCIS (Netherlands)

    Golliard, J.; Belfroid, S.P.C.; Bendiksen, E.; Frimodt, C.

    2013-01-01

    Pipes for gas production and transport with a corrugated inner surface, as used in flexible pipes, can be subject to Flow-Induced Pulsations when the flow velocity is larger than a certain velocity. This onset velocity is dependent on the geometry of the corrugations, the operational conditions and

  19. Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break)

    International Nuclear Information System (INIS)

    Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1976-07-01

    Recorded test data are presented for Test S-29-1 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 0 F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

  20. LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The third OECD LOFT experiment was conducted on 14 July 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  1. Sampling based uncertainty analysis of 10% hot leg break LOCA in large scale test facility

    International Nuclear Information System (INIS)

    Sengupta, Samiran; Kraina, V.; Dubey, S. K.; Rao, R. S.; Gupta, S. K.

    2010-01-01

    Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between 5 th and 95 th percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure

  2. Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

  3. LOFT/L3-, Loss of Fluid Test, 7. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the seventh in the NRC L3 Series of small-break LOCA experiments. A 2.5-cm (10-in.) cold-leg non-communicative-break LOCA was simulated. The experiment was conducted on 20 June 1980

  4. LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The second OECD LOFT experiment was conducted on 23 June 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  5. Technical report on the fatigue crack Growth Benchmark based on CEA pipe bending tests

    International Nuclear Information System (INIS)

    2001-07-01

    In order to improve the estimation methods of surface crack propagation through the thickness of components, CEA has proposed a benchmark to members of the IAGE WG, sub-group on Integrity of metal components and structures. The subject is a simple configuration of a pipe containing an axisymmetric notch and submitted to a cyclic bending load. An experimental data-set form CEA was used to validate three issues in the topic of Leak Before Break. - Crack initiation, - Crack propagation through the thickness, - Crack penetration. All material and geometrical data which are necessary for the simulation were given in the proposal, including experimental results. Due to the peculiar complexity of the problem, it was decided to focus the work on methodologies comparison so as to allow participants to tune up parameters and adjust their models and tools. This report presents all estimations performed by the participants and collected by CEA. They are compared to the experimental results. An analysis of the used procedures is also proposed. This, associated with the study of the accuracy of different methodologies, leads to comments and recommendations on the analysis of fatigue crack growth. The participation in the first step was important: nine participants have proposed analyses, sometimes parametric analysis to estimate crack growth. Results sorted out three estimation methods groups that give results in accordance with experimental ones (these three groups are based on a strain range evaluation and the fatigue curve of the material): - The use of an elastic stress at the notch tip and a fatigue notch concentration factor to determine the strain range. - The use of a KI (or elastic F.E. calculation) and a Neuber rule for the estimation of the strain range at a characteristic distance from the crack tip. - The direct calculation of the strain range at the characteristic distance by an elastic plastic F.E. calculation. Only 4 participants have proposed an estimate of the

  6. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  7. TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

  8. Comparison of fracture toughness values from an IPIRG-1 large-scale pipe system test and C(T) specimens on wrought TP304 stainless steel

    International Nuclear Information System (INIS)

    Olson, R.J.; Scott, P.; Marschall, C.W.; Wilkowski, G.M.

    1994-01-01

    Within the First International Piping Integrity Research Group (IPIRG-1) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe, and the experiments were conducted at a pressure of 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. Three independent analyses were used to calculate the toughness directly from one of these pipe experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment to standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria

  9. Residual stress state in pipe cut ring specimens for fracture toughness testing

    Energy Technology Data Exchange (ETDEWEB)

    Damjanovic, Darko [J.J. Strossmayer Univ. of Osijek, Slavonski Brod (Croatia). Mechanical Engineering Faculty; Kozak, Drazan [Zagreb Univ. (Croatia). Dept. for Mechanical Design; Marsoner, Stefan [Materials Center, Leoben (Austria).; Gubeljak, Nenad [Maribor Univ. (Slovenia). Chair of Mechanics

    2017-07-01

    Thin-walled pipes are not suitable for measuring fracture toughness parameters of vital importance because longitudinal crack failure is the most common failure mode in pipes. This is due to the impossibility to manufacture standard specimens for measuring fracture toughness, such as SENB or CT specimens, from the thin wall of the pipe. Previous works noticed this problem, but until now, a good and convenient solution has not been found or developed. To overcome this problem, very good alternative solution was proposed, the so-called pipe ring notched bend specimen (PRNB) [1-5]. Until now, only the idealized geometry PRNB specimen is analyzed, i. e., a specimen which is not cut out from an actual pipe but produced from steel plate. Based on that, residual stresses are neglected along with the imperfections in geometry (elliptical and eccentricity). The aim of this research is to estimate the residual stress state(s) in real pipes used in the boiler industry produced by hot rolling technique. These types of pipes are delivered only in normalized condition, but not stress relieved. Therefore, there are residual stresses present due to the manufacturing technique, but also due to uneven cooling after the production process. Within this paper, residual stresses are estimated by three methods: the incremental hole drilling method (IHMD), X-ray diffraction (XRD) and the splitting method (SM). Knowing the residual stress state in the ring specimen, it is possible to assess their impact on fracture toughness measured on the corresponding PRNB specimen(s).

  10. Residual stress state in pipe cut ring specimens for fracture toughness testing

    International Nuclear Information System (INIS)

    Damjanovic, Darko; Kozak, Drazan; Marsoner, Stefan; Gubeljak, Nenad

    2017-01-01

    Thin-walled pipes are not suitable for measuring fracture toughness parameters of vital importance because longitudinal crack failure is the most common failure mode in pipes. This is due to the impossibility to manufacture standard specimens for measuring fracture toughness, such as SENB or CT specimens, from the thin wall of the pipe. Previous works noticed this problem, but until now, a good and convenient solution has not been found or developed. To overcome this problem, very good alternative solution was proposed, the so-called pipe ring notched bend specimen (PRNB) [1-5]. Until now, only the idealized geometry PRNB specimen is analyzed, i. e., a specimen which is not cut out from an actual pipe but produced from steel plate. Based on that, residual stresses are neglected along with the imperfections in geometry (elliptical and eccentricity). The aim of this research is to estimate the residual stress state(s) in real pipes used in the boiler industry produced by hot rolling technique. These types of pipes are delivered only in normalized condition, but not stress relieved. Therefore, there are residual stresses present due to the manufacturing technique, but also due to uneven cooling after the production process. Within this paper, residual stresses are estimated by three methods: the incremental hole drilling method (IHMD), X-ray diffraction (XRD) and the splitting method (SM). Knowing the residual stress state in the ring specimen, it is possible to assess their impact on fracture toughness measured on the corresponding PRNB specimen(s).

  11. Structural Dynamic Assessment of the GN2 Piping System for NASA's New and Powerful Reverberant Acoustic Test Facility

    Science.gov (United States)

    McNelis, Mark E.; Staab, Lucas D.; Akers, James C.; Hughes, WIlliam O.; Chang, Li, C.; Hozman, Aron D.; Henry, Michael W.

    2012-01-01

    The National Aeronautics and Space Administration (NASA) Glenn Research Center (GRC) has led the design and build of the new world-class vibroacoustic test capabilities at the NASA GRC's Plum Brook Station in Sandusky, Ohio, USA from 2007-2011. SAIC-Benham has completed construction of a new reverberant acoustic test facility to support the future testing needs of NASA's space exploration program and commercial customers. The large Reverberant Acoustic Test Facility (RATF) is approximately 101,000 cu ft in volume and was designed to operate at a maximum empty chamber acoustic overall sound pressure level (OASPL) of 163 dB. This combination of size and acoustic power is unprecedented amongst the world's known active reverberant acoustic test facilities. Initial checkout acoustic testing was performed on March 2011 by SAIC-Benham at test levels up to 161 dB OASPL. During testing, several branches of the gaseous nitrogen (GN2) piping system, which supply the fluid to the noise generating acoustic modulators, failed at their "t-junctions" connecting the 12 inch supply line to their respective 4 inch branch lines. The problem was initially detected when the oxygen sensors in the horn room indicated a lower than expected oxygen level from which was inferred GN2 leaks in the piping system. In subsequent follow up inspections, cracks were identified in the failed "t-junction" connections through non-destructive evaluation testing . Through structural dynamic modeling of the piping system, the root cause of the "t-junction" connection failures was determined. The structural dynamic assessment identified several possible corrective design improvements to the horn room piping system. The effectiveness of the chosen design repairs were subsequently evaluated in September 2011 during acoustic verification testing to 161 dB OASPL.

  12. Gas dynamic design of the pipe line compressor with 90% efficiency. Model test approval

    Science.gov (United States)

    Galerkin, Y.; Rekstin, A.; Soldatova, K.

    2015-08-01

    Gas dynamic design of the pipe line compressor 32 MW was made for PAO SMPO (Sumy, Ukraine). The technical specification requires compressor efficiency of 90%. The customer offered favorable scheme - single-stage design with console impeller and axial inlet. The authors used the standard optimization methodology of 2D impellers. The original methodology of internal scroll profiling was used to minimize efficiency losses. Radically improved 5th version of the Universal modeling method computer programs was used for precise calculation of expected performances. The customer fulfilled model tests in a 1:2 scale. Tests confirmed the calculated parameters at the design point (maximum efficiency of 90%) and in the whole range of flow rates. As far as the authors know none of compressors have achieved such efficiency. The principles and methods of gas-dynamic design are presented below. The data of the 32 MW compressor presented by the customer in their report at the 16th International Compressor conference (September 2014, Saint- Petersburg) and later transferred to the authors.

  13. ISP-50 Specifications for a Direct Vessel Injection Line Break Test with the ATLAS

    International Nuclear Information System (INIS)

    Choi, Ki Yong; Baek, Won Pil; Kim, Yeon Sik; Park, Hyun Sik; Cho, Seok; Kang, Kyoung Ho; Choi, Nam Hyun; Min, Kyoung Ho

    2009-06-01

    An OECD/NEA International Standard Problem Exercise (ISP) focussing on a DVI line break simulation result with the ATLAS was approved by the NEA Committee on the Safety of Nuclear Installation (CSNI) meeting in December 2008 and was numbered by ISP-50. The ISP-50 program will be operated by an operating agency, KAERI for three years starting from the physical year 2009. Fourteen international organizations confirmed their participation in the ISP-50, including NRC (USA), JAEA, JNES (Japan), GRS (Germany), KFKI-AEKI (Hungary), EDO Gidropress (Russia), VTT, Fortum (Finland), NRI (Czech Republic), Univ. of Pisa (Italy), KINS, KNF, KOPEC, and KAERI (Korea). In addition, KTH in Sweden and HSE in UK are considering late participation. Recently, NPIC and CIAE in China hope to join the ISP-50. As for the safety analysis codes, nine codes are expected to be used for the ISP-50: MARS-3D, RELAP5- 3D, RELAP5, TRACE, CATHARE, APROS, ATHELET, TRAP, and KORSAR. It is the first ISP exercise in Korea in which a domestic test facility is utilized by international nuclear society and this exercise will contribute to extending our physical understanding on thermal hydraulic phenomena during the DVI line break accidents and to verifying the best-estimate thermal-hydraulic safety analysis codes. This report was prepared to define technical specifications of the ISP-50 exercise according the guideline provided by OECD/CSNI. It includes general objectives, phases, deliverables to participants, parameters required for comparison and the time table

  14. Modelling of Aquitaine II pipe whipping test with EUROPLEXUS fast dynamics code

    International Nuclear Information System (INIS)

    Potapov, S.

    2003-01-01

    To validate the modelling of multi-physics phenomena with EUROPLEXUS code we considered a pipe whipping problem occurring in thermal hydraulic conditions of a Loss of Coolant Accident in PWR primary circuit. Two numerical fluid-structure interaction (FSI) models, a simplified 'pipe-like' model and a mixed 1D/3D model, were used to simulate both the conditioning phase and a phase of whipping. The results of calculations were compared with existing experimental data. Analysis of numerical results shows that both models give a good prediction of global behaviour of the coupled fluid-structure system, namely for pipe displacements and stresses in the pipe walls, as well as for pressure and velocity in the fluid. By comparison with experimental data, we show that only the mixed EUROPLEXUS model, where the pipe elbow is discretized with shells, allows us to estimate correctly the time history and maximum value of the contact force between the pipe and the obstacle. The 1D model with reduced kinematics (rigid cross section hypothesis) does not allow the correct detection of contact phenomenon. This study shows that the use of mixed numerical models containing simplified and totally 3D parts duly interconnected allows a very efficient and CPU inexpensive numerical analysis which is able to take into account different global and local physical phenomena. (author)

  15. Recirculation pump suction line 2.8% break integral test at ROSA-III with HPCS failure, RUN 984

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Anoda, Yoshinari; Tasaka, Kanji; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Shiba, Masayoshi

    1984-06-01

    This report presents the experimental data of 2.8% suction line break test RUN 984 at ROSA-III, which was conducted as one of counterpart tests to FIST program sponsored by GE, EPRI and USNRC. The similarity study between the ROSA-III and FIST tests is on the way. The report also presents the information on the ROSA-III test facility, experiment results and the effects of the ADS flow rate and the MSIV trip level comparing with the previously conducted ROSA-III small break tests, RUNs 920 and 922. Major conclusions obtained are as follows. (1) Change of the MSIV trip level from L2 to L1 gives delay of MSIV closure and longer actuation of pressure control system in a small break LOCA. (2) Larger ADS flow gives faster depressurization rate and earlier ECCS actuation, which results in shorter fuel rod dryout period and lower PCT. (author)

  16. Automated Portfolio Optimization Based on a New Test for Structural Breaks

    Directory of Open Access Journals (Sweden)

    Tobias Berens

    2014-04-01

    Full Text Available We present a completely automated optimization strategy which combines the classical Markowitz mean-variance portfolio theory with a recently proposed test for structural breaks in covariance matrices. With respect to equity portfolios, global minimum-variance optimizations, which base solely on the covariance matrix, yield considerable results in previous studies. However, financial assets cannot be assumed to have a constant covariance matrix over longer periods of time. Hence, we estimate the covariance matrix of the assets by respecting potential change points. The resulting approach resolves the issue of determining a sample for parameter estimation. Moreover, we investigate if this approach is also appropriate for timing the reoptimizations. Finally, we apply the approach to two datasets and compare the results to relevant benchmark techniques by means of an out-of-sample study. It is shown that the new approach outperforms equally weighted portfolios and plain minimum-variance portfolios on average.

  17. Linkages among commodity futures prices in the recent financial crisis: An application of cointegration tests with a structural break

    Directory of Open Access Journals (Sweden)

    Yoichi Tsuchiya

    2015-12-01

    Full Text Available In this study, we investigate the existence of long-term co-movements among the prices of commodity futures contracts. We use a cointegration test, which accounts for the presence of a structural break. We show that while there is a long-term relationship among agricultural and among non-agricultural commodity futures prices when a structural break is taken into account, there is no such relationship without allowing for a structural break. We also show that these break points, in fact, occur a few months before the recent global financial crisis. Although the previous literature broadly casts doubt on such price co-movements, our results confirm that market performance improved during the sample period.

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

    International Nuclear Information System (INIS)

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

    2003-12-01

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

  19. Standard Test Method for Hydrophobic Surface Films by the Water-Break Test

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This test method covers the detection of the presence of hydrophobic (nonwetting) films on surfaces and the presence of hydrophobic organic materials in processing ambients. When properly conducted, the test will enable detection of molecular layers of hydrophobic organic contaminants. On very rough or porous surfaces, the sensitivity of the test may be significantly decreased. 1.2 The values stated in SI units are to be regarded as the standard. The inch-pound values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  20. Breaking HIV News to Clients: SPIKES Strategy in Post-Test Counseling Session

    Directory of Open Access Journals (Sweden)

    Hamid Emadi-Koochak

    2016-05-01

    Full Text Available Breaking bad news is one of the most burdensome tasks physicians face in their everyday practice. It becomes even more challenging in the context of HIV+ patients because of stigma and discrimination. The aim of the current study is to evaluate the quality of giving HIV seroconversion news according to SPIKES protocol. Numbers of 154 consecutive HIV+ patients from Imam Khomeini Hospital testing and counseling center were enrolled in this study. Patients were inquired about how they were given the HIV news and whether or not they received pre- and post-test counseling sessions. Around 51% of them were men, 80% had high school education, and 56% were employed. Regarding marital status, 32% were single, and 52% were married at the time of the interview. Among them, 31% had received the HIV news in a counseling center, and only 29% had pre-test counseling. SPIKES criteria were significantly met when the HIV news was given in an HIV counseling and testing center (P.value<0.05. Low coverage of HIV counseling services was observed in the study. SPIKES criteria were significantly met when the HIV seroconversion news was given in a counseling center. The need to further train staff to deliver HIV news seems a priority in the field of HIV care and treatment.

  1. 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

  2. Qualification methodologies for mechanical component, I and C, piping using test lab

    International Nuclear Information System (INIS)

    Ichikawa, Toshio

    2001-01-01

    There are many methods of verifying the intensity of a structure, a function, a vibration characteristics, etc. The seismic test which verifies the function during the earthquake of a components simple substance (seismic test which checks durability according to components types). How to verify the analysis technique by the scale model and to check the intensity of plant operating conditions by the scale model results. The model of the same size as the actual plant is created and there is a method of verifying intensity and the function directly. A seismic test is restrained by the frequency of an evaluation objective, and the capability of actuator equipment, and is carried out. Moreover, otherwise, restrictions are the size of a table, actuation power, environment, etc. Here, further examples are introduced, such as evaluation by the examination that combined analysis, experimental test use and analysis, and the experimental test, and the method of proving only by test, and have the seismic check method by test learned in this lecture. Typical examples are explained. Based on the seismic test result carried out with experimental research equipment, how to verify that the required function to components, such as a structure of reactor internals, is maintained at the time of an earthquake is explained. In this case, differences of the simulation environment of the model in. a test, earthquake conditions simulated by shaker table of test conditions and actual plant conditions are important for the evaluation method determination. In nuclear equipment, there is what is required to achieve the static function to hold pressure boundary to the high temperature inside apparatus piping - high-pressure flow, and dynamic functions, such as insertion of a valve, a pump, and a control rod. Moreover, in order to maintain and carry out the safe stop of the safe operation, there is I and C for controlling - supervising these components. In order for this functional maintenance

  3. Development of Test Rig for Robotization of Mining Technological Processes - Oversized Rock Breaking Process Case

    Science.gov (United States)

    Pawel, Stefaniak; Jacek, Wodecki; Jakubiak, Janusz; Zimroz, Radoslaw

    2017-12-01

    Production chain (PCh) in underground copper ore mine consists of several subprocesses. From our perspective implementation of so called ZEPA approach (Zero Entry Production Area) might be very interesting [16]. In practice, it leads to automation/robotization of subprocesses in production area. In this paper was investigated a specific part of PCh i.e. a place when cyclic transport by LHDs is replaced with continuous transport by conveying system. Such place is called dumping point. The objective of dumping points with screen is primary classification of the material (into coarse and fine material) and breaking oversized rocks with hydraulic hammer. Current challenges for the underground mining include e.g. safety improvement as well as production optimization related to bottlenecks, stoppages and operational efficiency of the machines. As a first step, remote control of the hydraulic hammer has been introduced, which not only transferred the operator to safe workplace, but also allowed for more comfortable work environment and control over multiple technical objects by a single person. Today literature analysis shows that current mining industry around the world is oriented to automation and robotization of mining processes and reveals technological readiness for 4th industrial revolution. The paper is focused on preliminary analysis of possibilities for the use of the robotic system to rock-breaking process. Prototype test rig has been proposed and experimental works have been carried out. Automatic algorithms for detection of oversized rocks, crushing them as well as sweeping and loosening of material have been formulated. Obviously many simplifications have been assumed. Some near future works have been proposed.

  4. Optimized Heat Pipe Backup Cooling System Tested with a Stirling Convertor

    Science.gov (United States)

    Schwendeman, Carl L.; Tarau, Calin; Schifer, Nicholas A.; Anderson, William G.; Garner, Scott

    2016-01-01

    In a Stirling Radioisotope Power System (RPS), heat must be continuously removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS at the cost of an early termination of the mission. An alkali-metal variable conductance heat pipe (VCHP) can be used to passively allow multiple stops and restarts of the Stirling convertor by bypassing the heat during stops. In a previous NASA Small Business Innovation Research (SBIR) Program, Advanced Cooling Technologies, Inc. (ACT) developed a series of sodium VCHPs as backup cooling systems for the Stirling RPS. In 2012, one of these VCHPs was successfully tested at NASA Glenn Research Center with a Stirling convertor as an Advanced Stirling Radioisotope Generator (ASRG) backup cooling system. The prototype; however, was not optimized and did not reflect the final heat rejection path. ACT through further funding has developed a semioptimized prototype with the finalized heat path for testing at Glenn with a Stirling convertor. The semioptimized system features a two-phase radiator and is significantly smaller and lighter than the prior prototype to reflect a higher level of flight readiness. The VCHP is designed to activate and remove heat from the GPHS during stoppage with a small temperature increase from the nominal vapor temperature. This small temperature increase from nominal is low enough to avoid risking standard ASRG operation and spoiling of the multilayer insulation (MLI). The VCHP passively allows the Stirling convertor to be turned off multiple times during a mission with potentially unlimited off durations. Having the ability to turn the Stirling off allows for the Stirling to be reset and reduces vibrations on the platform during sensitive measurements or

  5. LOFT/LP-02-6, Loss of Fluid Test, 1. OECD Large Break Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The fourth OECD LOFT experiment was conducted on 3 October 1983. This was the first OECD LOFT large break experiment. The initial and boundary conditions were chosen to be representative of USNRC licensing limits for commercial PWRs. This included loss of off-site power coincident with LOCA initiation. This experiment included the first use in LOFT of pressurized fuel rods in the center bundle. The experiment was initiated by opening the quick-opening blow-down valves in the broken hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  6. Qualification of new design of flexible pipe against singing: testing at multiple scales

    NARCIS (Netherlands)

    Golliard, J.; Lunde, K.; Vijlbrief, O.

    2016-01-01

    Flexible pipes for production of oil and gas typically present a corrugated inner surface. This has been identified as the cause of "singing risers": Flow-Induced Pulsations due to the interaction of sound waves with the shear layers at the small cavities present at each of the multiple

  7. 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

  8. Development and testing of passive autocatalytic recombiners cooled by heat pipes; Entwicklung und Erprobung mittels Heatpipe gekuehlter katalytischer Rekombinatoren

    Energy Technology Data Exchange (ETDEWEB)

    Granzow, Christoph

    2012-11-26

    A severe accident in a nuclear power plant (NPP) can lead to core damage in conjunction with the release of large amounts of hydrogen. As hydrogen mitigation measure, passive autocatalytic recombiners (PARs) are used in today's pressurized water reactors. PARs recombine hydrogen and oxygen contained in the air to steam. The heat from this exothermic reaction causes the catalyst and its surroundings to heat up. If parts of the PAR heat up above the ignition temperature of the gas mixture, a spontaneous deflagration or detonation can occur. The aim of this work is the prevention of such high temperatures by means of passive cooling of the catalyst with heat pipes. Heat pipes are completely passive heat exchanger with a very high effective thermal conductivity. For a deeper understanding of the reaction kinetics at lower temperatures, single catalytic coated heat pipes are studied in a flow reactor. The development of a modular small-scale PAR model is then based on a test series with cooled catalyst sheets. Finally, the PAR model is tested inside a pressure vessel under boundary conditions similar to a real NPP. The experiments show, that the temperatures of the cooled catalytic sheets stay significantly below the temperature of the uncooled sheets and below the ignition temperature of the gas mixture under any set boundary conditions, although no significant reduction of the conversion efficiency can be observed. As a last point, a mathematical model of the reaction kinetics of the recombination process as well as a model of the fluid dynamic and thermohydraulic processes in a heat pipe are developed with the data obtained from the experiments.

  9. Visualization test using piping group mock up specimen for evaluation of wastage phenomena in steam generator for FBR

    International Nuclear Information System (INIS)

    Kato, Keisuke; Yoshida, Atsuro; Arae, Kunihiko; Narabayashi, Tadashi; Ohshima, Hiroyuki; Kurihara, Akikazu

    2012-01-01

    There is a need for quantitative evaluation of wastage phenomena in steam generator for FBR. We focused attention on liquid droplet impingement erosion (LDIE) in wastage phenomena and performed basic study with piping group mock up specimen for quantitative evaluation of LDIE. First, we did visualization test of high pressure and high speed jet into the water. Test section mock up the crack of heat exchanger tube and neighboring heat exchanger tubes. We did the test under the following test conditions. Upstream pressure is 0.3MPa, vapor temperature is 300K, crack width is 0.1mm, and crack length is 40mm. (crack diameter is 0.2mm) Second, we did pressure and temperature measurement test in the same test conditions as before. We evaluated jet behavior at test section by those two tests. In addition, we did two phase flow analysis of the jet with TRAC code. (author)

  10. Modelling and testing the molecular mechanism of radiation-induced chromatid breaks

    International Nuclear Information System (INIS)

    Bryant, P. E.

    2001-01-01

    Chromatid breaks induced by ionizing radiation in the G2 phase of the cell cycle are considered as markers of individual human radiosensitivity and may indicate the presence of low-penetrance genes regulating susceptibility to breast and other cancers). Together with our own study of Scottish (Tayside) breast cancer patients and a group of normal controls these studies show an overall 10-fold variation in chromatid break frequency (the parameter defining individual chromosomal 'radiosensitivity'). Thus, an understanding of the mechanisms and genes involved in determining these widely different responses should help to clarify the reasons for individual radiosensitivity and may lead us to identify key genes involved in cancer susceptibility. The presence of colour-switches at around 16% of chromatid break points (detected in harlequin-stained chromosomes) indicates that this type of chromatid break is formed by a chromatin rearrangement involving one or more large chromatin domains of the order of 3 - 5 Mbp, possibly representing transcription 'factories'. The signal model of chromatid breaks assumes that all chromatid breaks are the result of chromatin rearrangements, and that the initiating DNA double-strand break (dsb) is itself not involved in the rearrangement but signals its presence (possibly via ATM protein or DNAPK) leading to the initiation of the chromatin rearrangement. Experimental evidence from radiosensitive cell lines (e.g. human AT and hamster irs2) and with the nucleoside analogue araA (9-β-D-arabinofuranosyladenine) demonstrates the lack of correspondence between the rejoining kinetics of dsb and that of disappearance of chromatid breaks, thus supporting the signal model. Coupled with the linear induction of chromatid breaks with dose in both human and rodent cell lines of various types, and the production of chromatid breaks by single dsb in genetically engineered cell lines the classical 'breakage-first' model of chromatid breaks is no longer

  11. Application of ultrasonic testing technique to detect gas accumulation in important pipings for pressurized water reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Fushimi, Yasuyuki [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    Since 1988, the USNRC has pointed out that gas-binding events might occur at high head safety injection (HHSI) pumps of pressurized water reactors (PWRs). In Japanese PWR plants, corrective actions were taken in response to gas-binding events that occurred on HHSI pumps in the USA, so no gas accumulation event has been reported so far. However, when venting frequency is prolonged with operating cycle extension, the probability of gas accumulation in pipings may increase as in the USA. The purpose of this study was to establish a technique to identify gas accumulation and to measure the gas volume accurately. Taking dominant causes of the gas-binding events in the USA into consideration, we pointed out the following sections in the Japanese PWRs where gas srtipping and/or gas accumulation might occur: residual heat removal system pipings and charging/safety injection pump minimum flow line. Then an ultrasonic testing technique, adopted to identify gas accumulation in the USA, was applied to those sections of the typical Japanese PWR. Consequently, no gas accumulation was found in those pipings. (author)

  12. Eddy Current Testing with Giant Magnetoresistance (GMR) Sensors and a Pipe-Encircling Excitation for Evaluation of Corrosion under Insulation.

    Science.gov (United States)

    Bailey, Joseph; Long, Nicholas; Hunze, Arvid

    2017-09-28

    This work investigates an eddy current-based non-destructive testing (NDT) method to characterize corrosion of pipes under thermal insulation, one of the leading failure mechanisms for insulated pipe infrastructure. Artificial defects were machined into the pipe surface to simulate the effect of corrosion wall loss. We show that by using a giant magnetoresistance (GMR) sensor array and a high current (300 A), single sinusoidal low frequency (5-200 Hz) pipe-encircling excitation scheme it is possible to quantify wall loss defects without removing the insulation or weather shield. An analysis of the magnetic field distribution and induced currents was undertaken using the finite element method (FEM) and analytical calculations. Simple algorithms to remove spurious measured field variations not associated with defects were developed and applied. The influence of an aluminium weather shield with discontinuities and dents was ascertained and found to be small for excitation frequency values below 40 Hz. The signal dependence on the defect dimensions was analysed in detail. The excitation frequency at which the maximum field amplitude change occurred increased linearly with the depth of the defect by about 3 Hz/mm defect depth. The change in magnetic field amplitude due to defects for sensors aligned in the azimuthal and radial directions were measured and found to be linearly dependent on the defect volume between 4400-30,800 mm³ with 1.2 × 10 -3 -1.6 × 10 -3 µT/mm³. The results show that our approach is well suited for measuring wall loss defects similar to the defects from corrosion under insulation.

  13. 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.

  14. Simulation and analysis on ultrasonic testing for the cement grouting defects of the corrugated pipe

    Energy Technology Data Exchange (ETDEWEB)

    Qingbang, Han; Ling, Chen; Changping, Zhu [Changzhou Key Laboratory of Sensor Networks and Environmental Sensing, College of IOT, Hohai University Changzhou, Jiangsu, 213022 (China)

    2014-02-18

    The defects exist in the cement grouting process of prestressed corrugated pipe may directly impair the bridge safety. In this paper, sound fields propagation in concrete structures with corrugated pipes and the influence of various different defects are simulated and analyzed using finite element method. The simulation results demonstrate a much complex propagation characteristic due to multiple reflection, refraction and scattering, where the scattering signals caused by metal are very strong, while the signals scattered by an air bubble are weaker. The influence of defect both in time and frequency domain are found through deconvolution treatment. In the time domain, the deconvolution signals correspond to larger defect display a larger head wave amplitude and shorter arrive time than those of smaller defects; in the frequency domain, larger defect also shows a stronger amplitude, lower center frequency and lower cutoff frequency.

  15. Simulation and analysis on ultrasonic testing for the cement grouting defects of the corrugated pipe

    International Nuclear Information System (INIS)

    Qingbang, Han; Ling, Chen; Changping, Zhu

    2014-01-01

    The defects exist in the cement grouting process of prestressed corrugated pipe may directly impair the bridge safety. In this paper, sound fields propagation in concrete structures with corrugated pipes and the influence of various different defects are simulated and analyzed using finite element method. The simulation results demonstrate a much complex propagation characteristic due to multiple reflection, refraction and scattering, where the scattering signals caused by metal are very strong, while the signals scattered by an air bubble are weaker. The influence of defect both in time and frequency domain are found through deconvolution treatment. In the time domain, the deconvolution signals correspond to larger defect display a larger head wave amplitude and shorter arrive time than those of smaller defects; in the frequency domain, larger defect also shows a stronger amplitude, lower center frequency and lower cutoff frequency

  16. 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

  17. A method for predicting the fatigue life of pre-corroded 2024-T3 aluminum from breaking load tests

    Science.gov (United States)

    Gruenberg, Karl Martin

    Characterization of material properties is necessary for design purposes and has been a topic of research for many years. Over the last several decades, much progress has been made in identifying metrics to describe fracture mechanics properties and developing procedures to measure the appropriate values. However, in the context of design, there has not been as much success in quantifying the susceptibility of a material to corrosion damage and its subsequent impact on material behavior in the framework of fracture mechanics. A natural next step in understanding the effects of corrosion damage was to develop a link between standard material test procedures and fatigue life in the presence of corrosion. Simply stated, the goal of this investigation was to formulate a cheaper and quicker method for assessing the consequences of corrosion on remaining fatigue life. For this study, breaking load specimens and fatigue specimens of a single nominal gage (0.063″) of aluminum alloy 2024-T3 were exposed to three levels of corrosion. The breaking load specimens were taken from three different material lots, and the fatigue tests were carried out at three stress levels. All failed specimens, both breaking load and fatigue specimens, were examined to characterize the damage state(s) and failure mechanism(s). Correlations between breaking load results and fatigue life results in the presence of corrosion damage were developed using a fracture mechanics foundation and the observed mechanisms of failure. Where breaking load tests showed a decrease in strength due to increased corrosion exposure, the corresponding set of fatigue tests showed a decrease in life. And where breaking load tests from different specimen orientations exhibited similar levels of strength, the corresponding set of fatigue specimens showed similar lives. The spread from shortest to longest fatigue lives among the different corrosion conditions decreased at the higher stress levels. Life predictions based

  18. 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)

  19. 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

  20. 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...

  1. Evaluation of Effect by Internal Flow on Ultrasonic Testing Flaw Sizing in Piping

    International Nuclear Information System (INIS)

    Lee, Jeong Seok; Yoon, Byung Sik; Kim, Yong Sik

    2013-01-01

    In this study, the ultrasonic amplitude difference between air filled and water filled piping in nuclear power plant is compared by modeling approach. In this study, ultrasonic amplitude differences between air and water filled pipe are evaluated by modeling approach. Consequently, we propose the following results. The ultrasonic amplitude difference between air and water filled condition is measured by lower than 1 dB in modeling calculation. The flaw length sizing error between air and water filled condition shows same results based on 12 dB drop method even thought the amplitude difference is 1 dB. Most of the piping welds in nuclear power plants are inspected periodically using ultrasonic techniques to detect service-induced flaws such as IGSCC cracking. The inspection results provide information such as location, maximum amplitude response, ultrasonic length, height and finally the nature or flaw pattern. The founded flaw in ultrasonic inspection is accepted or rejected based on these information. Specially, the amplitude of flaw response is very important to estimate the flaw size. Currently the ultrasonic inspections in nuclear power plant components are performed by specific inspection procedure which describing inspection technique include inspection system, calibration methodology and flaw characterizing methodology. To perform ultrasonic inspection during in-service inspection, reference gain should be established before starting ultrasonic inspection by requirement of ASME code. This reference gain used as basic criteria to evaluate flaw sizing. Sometimes, a little difference in establishing reference gain between calibration and field condition can lead to deviation in flaw sizing. Due to this difference, the inspection result may cause flaw sizing error

  2. BWR 200 % recirculation pump suction line break LOCA tests, RUNs 942 and 943 at ROSA-III without HPCS

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Koizumi, Yasuo

    1986-03-01

    This report presents the experimental results of RUNs 942 and 943 in ROSA-III program, which are 200 % recirculation pump suction line break LOCA tests with assumption of HPCS failure. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, ECCS's, and steam and feedwater systems. Effects of initial core void distribution and other fluid conditions on overall LOCA phenomena with special interest on transient core cooling phenomena were investigated by comparing the present test results with those of RUN 926, a 200 % suction line break test with standard initial fluid conditions. The initial core outlet quality was changed between 5 % and 43 %. As conclusions, (1) the initial lower core flow and higher void fraction affected significantly the core cooling conditions and resulted in earlier and higher PCT. (2) The lower plenum flashing temporarily contributed to cool down the core. (3) Flashing of remained hot water in the feedwater line affected slightly the pressure response and delayed the actuation of LPCI by 11 seconds. (4) The whole core was completely cooled down within 104 seconds after the LPCI actuation in these large break tests. (author)

  3. Studies on split heat pipe type adsorption ice-making test unit for fishing boats: Choice of heat pipe medium and experiments under unsteady heating sources

    International Nuclear Information System (INIS)

    Wang, L.W.; Wang, R.Z.; Lu, Z.S.; Chen, C.J.

    2006-01-01

    The split heat pipe type compound adsorption ice maker for fishing boats not only has the advantage of large volume cooling density but also has the advantage of less power consumption and high heat transfer performance. The available heat pipe media for the split heat pipe type compound adsorption ice maker, which are methanol, acetone and water are studied and compared in this paper, and the heat pipe medium of water shows the better performance for the reason of its stable heating and cooling process and high heat transfer performance. Considering the waste heat recovered from the diesel engine on fishing boats varies when the velocity of the fishing boat changes, the refrigeration performances at the condition of different values of heating power are studied while water is used as the heat pipe medium. Results show that the cooling power, as while as COP and SCP decrease when the heating power decreases. The highest COP and SCP are 0.41 and 731 W/kg, respectively, at the highest heating power of 4.2 kW, and the values decrease by 22% and 33%, respectively, when the heating power decreases by 15%. The values decrease by 32% and 51%, respectively, when the heating power decreases by 30%. The performance of the adsorption ice maker for the fishing boat with the 6160A type diesel engine is estimated, and the results show that the cooling power and ice productivity are as high as 5.44 kW and 1032 kg ice per day, respectively, even if the recovered waste heat decreases by 30% compared with the normal value. It can satisfy the ice requirements of such a fishing boat

  4. 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.

  5. Medical students' attitudes towards breaking bad news: an empirical test of the World Health Organization model.

    NARCIS (Netherlands)

    Valck, C. de; Bensing, J.; Bruynooghe, R.

    2001-01-01

    The literature regarding breaking bad news distinguishes three disclosure models: non-disclosure, full-disclosure and individualized disclosure. In this study, we investigated the relations between attitudes regarding disclosure of bad news and global professional attitudes regarding medical care in

  6. CERN Control Centre (CCC) - Collision tests at record-breaking energy of 13 TeV

    CERN Multimedia

    Brice, Maximilien

    2015-01-01

    during the night of 20 May, protons collided in the Large Hadron Collider (LHC) at the record-breaking energy of 13 TeV for the first time. These pictures show the LHC operations team on the morning of 21 of May.

  7. An Eddy Current Testing Platform System for Pipe Defect Inspection Based on an Optimized Eddy Current Technique Probe Design

    Science.gov (United States)

    Rifai, Damhuji; Abdalla, Ahmed N.; Razali, Ramdan; Ali, Kharudin; Faraj, Moneer A.

    2017-01-01

    The use of the eddy current technique (ECT) for the non-destructive testing of conducting materials has become increasingly important in the past few years. The use of the non-destructive ECT plays a key role in the ensuring the safety and integrity of the large industrial structures such as oil and gas pipelines. This paper introduce a novel ECT probe design integrated with the distributed ECT inspection system (DSECT) use for crack inspection on inner ferromagnetic pipes. The system consists of an array of giant magneto-resistive (GMR) sensors, a pneumatic system, a rotating magnetic field excitation source and a host PC acting as the data analysis center. Probe design parameters, namely probe diameter, an excitation coil and the number of GMR sensors in the array sensor is optimized using numerical optimization based on the desirability approach. The main benefits of DSECT can be seen in terms of its modularity and flexibility for the use of different types of magnetic transducers/sensors, and signals of a different nature with either digital or analog outputs, making it suited for the ECT probe design using an array of GMR magnetic sensors. A real-time application of the DSECT distributed system for ECT inspection can be exploited for the inspection of 70 mm carbon steel pipe. In order to predict the axial and circumference defect detection, a mathematical model is developed based on the technique known as response surface methodology (RSM). The inspection results of a carbon steel pipe sample with artificial defects indicate that the system design is highly efficient. PMID:28335399

  8. The development and testing of the thermal break divertor monoblock target design delivering 20 MW m-2 heat load capability

    Science.gov (United States)

    Fursdon, M.; Barrett, T.; Domptail, F.; Evans, Ll M.; Luzginova, N.; Greuner, N. H.; You, J.-H.; Li, M.; Richou, M.; Gallay, F.; Visca, E.

    2017-12-01

    The design and development of a novel plasma facing component (for fusion power plants) is described. The component uses the existing ‘monoblock’ construction which consists of a tungsten ‘block’ joined via a copper interlayer to a through CuCrZr cooling pipe. In the new concept the interlayer stiffness and conductivity properties are tuned so that stress in the principal structural element of the component (the cooling pipe) is reduced. Following initial trials with off-the-shelf materials, the concept was realized by machined features in an otherwise solid copper interlayer. The shape and distribution of the features were tuned by finite element analyses subject to ITER structural design criterion in-vessel components (SDC-IC) design rules. Proof of concept mock-ups were manufactured using a two stage brazing process verified by tomography and micrographic inspection. Full assemblies were inspected using ultrasound and thermographic (SATIR) test methods at ENEA and CEA respectively. High heat flux tests using IPP’s GLADIS facility showed that 200 cycles at 20 MW m-2 and five cycles at 25 MW m-2 could be sustained without apparent component damage. Further testing and component development is planned.

  9. The development and testing of the thermal break divertor monoblock target design delivering 20 MW m−2 heat load capability

    International Nuclear Information System (INIS)

    Fursdon, M; Barrett, T; Domptail, F; Evans, Ll M; Luzginova, N; Greuner, N H; You, J-H; Li, M; Richou, M; Gallay, F; Visca, E

    2017-01-01

    The design and development of a novel plasma facing component (for fusion power plants) is described. The component uses the existing ‘monoblock’ construction which consists of a tungsten ‘block’ joined via a copper interlayer to a through CuCrZr cooling pipe. In the new concept the interlayer stiffness and conductivity properties are tuned so that stress in the principal structural element of the component (the cooling pipe) is reduced. Following initial trials with off-the-shelf materials, the concept was realized by machined features in an otherwise solid copper interlayer. The shape and distribution of the features were tuned by finite element analyses subject to ITER structural design criterion in-vessel components (SDC-IC) design rules. Proof of concept mock-ups were manufactured using a two stage brazing process verified by tomography and micrographic inspection. Full assemblies were inspected using ultrasound and thermographic (SATIR) test methods at ENEA and CEA respectively. High heat flux tests using IPP’s GLADIS facility showed that 200 cycles at 20 MW m −2 and five cycles at 25 MW m −2 could be sustained without apparent component damage. Further testing and component development is planned. (paper)

  10. Mechanized radiation testing of austenitic pipe welds. Testing of media filled pipes and determination of the flaw depth by tomosynthesis; Mechanisierte Durchstrahlungspruefung von Rundschweissnaehten. Pruefung mediengefuellter Rohrleitungen und Tiefenlagenbestimmung durch Tomosynthese

    Energy Technology Data Exchange (ETDEWEB)

    Ewert, U.; Redmer, B. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany); Mueller, J. [COMPRA GmbH, Frechen (Germany); Trobitz, M. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH, Gundremmingen (Germany); Baranov, V.A. [Institute for Introscopy, Tomsk (Russian Federation)

    1999-08-01

    A compact detection system was built for multi-angle inspection of pipes, consisting of a high-sensitivity radiometric line scanner and an ultrasonic manipulator. Improved flaw imaging quality is achieved with this system as compared to film radiography. Measurements have been carried out on site in a nuclear power plant and in a laboratory. Better flaw imaging quality was also achieved in the testing of water-filled pipes. Non-linear tomosynthesis was applied for processing and interpretation of measured data. The system delivers considerably better images of planary materials inhomogeneitites, (such as cracks and lack-of-bond defects). (orig./CB) [Deutsch] Eine hoch empfindliche radiometrische Zeilenkamera wurde mit einem Ultraschall-Manipulator zu einem Gesamtsystem aufgebaut und fuer Mehrwinkel-Inspektionen von Rohrleitungen angewandt. Bei der Inspektion von Rundschweissnaehten an Rohren mit ca. 8... 20 mm Wanddicke wurde eine Verbesserung der Bildqualitaet im Vergleich zur Filmradiographie erreicht. Diese Messungen wurden in einem Kernkraftwerk unter Vor-Ort-Bedingungen sowie im Labor ausgefuehrt. Ein signifikantes Ansteigen der Bildqualitaet wurde auch bei der Pruefung von wassergefuellten Rohren erzielt. Methoden der nicht-linearen Tomosynthese wurden fuer die Verarbeitung und Interpretation der gemessenen Projektionsdaten genutzt. Das entwickelte System gestattet eine erhebliche Verbesserung der Anzeige von planaren Materialinhomogenitaeten (z.B. Risse und Bindefehler). (orig./DGE)

  11. The Cryogenic Test Bed experiments: Cryogenic heat pipe flight experiment CRYOHP (STS-53). Cryogenic two phase flight experiment CRYOTP (STS-62). Cryogenic flexible diode flight experiment CRYOFD

    Science.gov (United States)

    Thienel, Lee; Stouffer, Chuck

    1995-09-01

    This paper presents an overview of the Cryogenic Test Bed (CTB) experiments including experiment results, integration techniques used, and lessons learned during integration, test and flight phases of the Cryogenic Heat Pipe Flight Experiment (STS-53) and the Cryogenic Two Phase Flight Experiment (OAST-2, STS-62). We will also discuss the Cryogenic Flexible Diode Heat Pipe (CRYOFD) experiment which will fly in the 1996/97 time frame and the fourth flight of the CTB which will fly in the 1997/98 time frame. The two missions tested two oxygen axially grooved heat pipes, a nitrogen fibrous wick heat pipe and a 2-methylpentane phase change material thermal storage unit. Techniques were found for solving problems with vibration from the cryo-collers transmitted through the compressors and the cold heads, and mounting the heat pipe without introducing parasitic heat leaks. A thermally conductive interface material was selected that would meet the requirements and perform over the temperature range of 55 to 300 K. Problems are discussed with the bi-metallic thermostats used for heater circuit protection and the S-Glass suspension straps originally used to secure the BETSU PCM in the CRYOTP mission. Flight results will be compared to 1-g test results and differences will be discussed.

  12. Statistical reliability assessment of UT round-robin test data for piping welds

    International Nuclear Information System (INIS)

    Kim, H.M.; Park, I.K.; Park, U.S.; Park, Y.W.; Kang, S.C.; Lee, J.H.

    2004-01-01

    Ultrasonic NDE is one of important technologies in the life-time maintenance of nuclear power plant. Ultrasonic inspection system is consisted of the operator, equipment and procedure. The reliability of ultrasonic inspection system is affected by its ability. The performance demonstration round robin was conducted to quantify the capability of ultrasonic inspection for in-service. Several teams employed procedures that met or exceeded with ASME sec. XI code requirements detected the piping of nuclear power plant with various cracks to evaluate the capability of detection and sizing. In this paper, the statistical reliability assessment of ultrasonic nondestructive inspection data using probability of detection (POD) is presented. The result of POD using logistic model was useful to the reliability assessment for the NDE hit or miss data. (orig.)

  13. Vibration monitoring of the primary piping systems during the hot functional tests of the Mulheim-Karlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1989-01-01

    During the hot functional tests of the Muelheim--Kaerlich first-of-a-kind plant, vibration measurements were made on the reactor pressure vessel and its' internals and on the primary piping system and main coolant pumps. This paper contains results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. Good agreement was found. In the course of these comparisons, information on the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained

  14. IN-SITU TEST EXPERIMENTAL RESEARCH ON LEAKAGE OF LARGE DIAMETER PRE-STRESSED CONCRETE CYLINDER PIPE (PCCP

    Directory of Open Access Journals (Sweden)

    Jianjun Luo

    2016-10-01

    Full Text Available In recent years, a big number of large diameter pre-stressed concrete cylinder pipe (PCCP lines have been applied to the Mid-route of the South-to-North Water Transfer Project. However, the leakage problem of PCCP causes annually heavy economic losses to our country. In such a context of situation, how to detect leaks rapidly and precisely after pipes appear cracks in water supply system has great significance. Based on the study and analysis of the characteristic structure of large diameter PCCP, a new leak detection system using fiber Bragg grating sensors, which can capture signals of water pressure change, is proposed. The feasibility, reliability and practicability of the system could be acceptable according to data achieved from in–situ tests. Moreover, the leak detection system can monitor in real-time of dynamic change of water pressure. The equations of the leakage quantity and water pressure have been presented in this paper, which can provide technical guidelines for large diameter PCCP lines maintenance.

  15. BWR 1 % main recirculation line break LOCA tests, RUNs 917 and 918, without HPCS at ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Okazaki, Motoaki; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1988-07-01

    In a case of small break loss-of-coolant accident (LOCA) at a boiling water reactor (BWR) system, it is important to lower the system pressure to cool down the reactor system by using either the high pressure core spray (HPCS) or the automatic depressurization system (ADS). The report presents characteristic test results of RUNs 918 and 917, which were performed at the rig-of-safety assessment (ROSA)-III program simulating a 1 % break BWR LOCA with an assumption of HPCS failure, and clarifies effects of the ADS delay time on a small break LOCA. The ROSA-III test facility simulates principal components of a BWR/6 system with volumetric scaling factor of 1/424. It is experimentally concluded that the ADS delay time shorter than 4 minutes results in a similar PCT as that in a standard case, in which the PCT is observed after actuation of the low pressure core spray (LPCS). And the ADS delay time longer than 4 minutes results in higher PCT than in the standard case. In the latter, the PCT depends on the ADS time, a 220 K higher PCT, for example, in a case of 10 minutes ADS delay compared with the standard case. (author) 52 refs. 299 figs

  16. Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su

    1996-10-01

    This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties

  17. Design of load-to-failure tests of high-voltage insulation breaks for ITER's cryogenic network

    CERN Document Server

    Langeslag, S A E; Aviles Santillana, I; Sgobba, S; Foussat, A

    2015-01-01

    The development of new generation superconducting magnets for fusion research, such as the ITER experiment, is largely based on coils wound with so-called cable-in-conduit conductors. The concept of the cable-in-conduit conductor is based on a direct cooling principle, by supercritical helium, flowing through the central region of the conductor, in close contact with the superconducting strands. Consequently, a direct connection exists between the electrically grounded helium coolant supply line and the highly energised magnet windings. Various insulated regions, constructed out of high-voltage insulation breaks, are put in place to isolate sectors with different electrical potential. In addition to high voltages and significant internal helium pressure, the insulation breaks will experience various mechanical forces resulting from differential thermal contraction phenomena and electro-magnetic loads. Special test equipment was designed, prepared and employed to assess the mechanical reliability of the insul...

  18. 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)

  19. Literature review and experimental investigation of heat pipes

    Science.gov (United States)

    Barsch, W. O.; Schoenhals, R. J.; Viskanta, R.; Winter, E. R. F.

    1971-01-01

    Tests on heat pipes determine operational limits, external boundary conditions, noncondensable gas effects, startup behavior, and geometric configurations. Experiment consists of design, construction, and testing of an apparatus for measuring wick properties, conventional heat pipes and coplanar heat pipes.

  20. 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)

  1. Testing the nature of the electroweak breaking from polarized e+e- beams

    International Nuclear Information System (INIS)

    Narison, S.; Wallet, J.C.

    1985-01-01

    We present a quite general expression of the longitudinal Asub(parallel) and the transverse Asub(perpendicular to) asymmetries for the single and pair production of electroweak (pseudo) scalars (Higgs, π tilde, sigma tilde, ssub(μ) ...) from polarized e + e - colliding beams via annihilation mechanism. We suggest hat the measurements of Asub(parallel) and Asub(perpendicular to) can reveal the nature of such spinless bosons and, then, the nature of the electroweak breaking, if these spinless bosons are produced at LEP and SLC energies. (orig.)

  2. BWR 2 % main recirculation line break LOCA tests RUNs 915 and 920 without HPCS in ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1987-03-01

    This report presents the experimental results of BWR LOCA integral tests, RUNs 915 and 920, which are performed in the ROSA-III program simulating 2 % main recirculation line break LOCA tests with and without pressure control system operation. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCS's, and steam and feedwater systems. The report presents (1) the experimental results of 2 % small break LOCA phanomena in the ROSA-III system and (2) the effects of the pressure control system on the LOCA phenomena. The pressure control system contributed to (A) prevent bulk flashing in the early blowdown phase, (B) early closure of MSIV by L2 level trip, (C) early actuation of ADS by L1 level trip. However, the core thermal responses of the two tests were similar because of the similar mass inventory in PV after the ADS actuation in both tests. (author)

  3. Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in 'like-new' conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1994-09-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the 'like-new' condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed

  4. Experiments on hydraulically-compensated Compressed Air Energy Storage (CAES) system using large-diameter vertical pipe two-phase flow test facility: test facility and test procedure

    International Nuclear Information System (INIS)

    Ohtsu, Iwao; Murata, Hideo; Kukita, Yutaka; Kumamaru, Hiroshige.

    1996-07-01

    JAERI, the University of Tokyo, the Central Research Institute of Electric Power Industry and Shimizu Corporation jointing performed and experimental study on two-phase flow in the hydraulically-compensated Compressed Air Energy Storage (CAES) system with a large-diameter vertical pipe two-phase flow test facility from 1993 to 1995. A hydraulically-compensated CAES system is a proposed, conceptual energy storage system where energy is stored in the form of compressed air in an underground cavern which is sealed by a deep (several hundred meters) water shaft. The shaft water head maintains a constant pressure in the cavern, of several mega Pascals, even during loading or unloading of the cavern with air. The dissolved air in the water, however, may create voids in the shaft when the water rises through the shaft during the loading, being forced by the air flow into the cavern. The voids may reduce the effective head of the shaft, and thus the seal may fail, if significant bubbling should occur in the shaft. This bubbling phenomenon (termed 'Champaign effect') and potential failure of the water seal ('blowout') are simulated in a scaled-height, scaled-diameter facility. Carbon dioxide is used to simulate high solubility of air in the full-height, full-pressure system. This report describes the expected and potential two-phase flow phenomena in a hydraulically-compensated CAES system, the test facility and the test procedure, a method to estimate quantities which are not directly measured by using measured quantities and hydrodynamic basic equations, and desirable additional instrumentation. (author)

  5. 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)

  6. Evaluation of J-R curve testing of nuclear piping materials using the direct current potential drop technique

    International Nuclear Information System (INIS)

    Hackett, E.M.; Kirk, M.T.; Hays, R.A.

    1986-08-01

    A method is described for developing J-R curves for nuclear piping materials using the DC Potential Drop (DCPD) technique. Experimental calibration curves were developed for both three point bend and compact specimen geometries using ASTM A106 steel, a type 304 stainless steel and a high strength aluminum alloy. These curves were fit with a power law expression over the range of crack extension encountered during J-R curve tests (0.6 a/W to 0.8 a/W). The calibration curves were insensitive to both material and sidegrooving and depended solely on specimen geometry and lead attachment points. Crack initiation in J-R curve tests using DCPD was determined by a deviation from a linear region on a plot of COD vs. DCPD. The validity of this criterion for ASTM A106 steel was determined by a series of multispecimen tests that bracketed the initiation region. A statistical differential slope procedure for determination of the crack initiation point is presented and discussed. J-R curve tests were performed on ASTM A106 steel and type 304 stainless steel using both the elastic compliance and DCPD techniques to assess R-curve comparability. J-R curves determined using the two approaches were found to be in good agreement for ASTM A106 steel. The applicability of the DCPD technique to type 304 stainless steel and high rate loading of ferromagnetic materials is discussed. 15 refs., 33 figs

  7. Vibration monitoring of the primary piping system during the hot functional tests of the Muelheim-Kaerlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1992-01-01

    During the hot functional tests of the Muelheim-Kaerlich plant, which was the first plant of its type, vibration measurements were made on the reactor pressure vessel and its internal parts and on the primary piping system and the main coolant pumps. This paper contains the results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement programs is to confirm that the components, which are of new structural design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. A good correspondence was found. In the course of these comparisons, information about the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained. The vibration of the main coolant pumps was also continuously monitored. The pump surveillance system for each pump includes two non-contacting displacement sensors for measuring the kinetic shaft orbit, as well as velocity sensors for recording the vibrational velocity of the pump motor housing. During the continuous monitoring, it was checked whether the signal amplitudes remained within the allowable limits. In addition the frequency content of the signals was determined periodically. In this way deviations could be detected immediately and be explained by means of subsequent correlation analysis. Thus amplitude changes resulting from resonance effects were identified. (orig.)

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  10. Development and Testing of a Linear Polarization Resistance Corrosion Rate Probe for Ductile Iron Pipe (Web Report 4361)

    Science.gov (United States)

    The North American water and wastewater community has hundreds of millions of feet of ductile iron pipe in service. Only a portion of the inventory has any form of external corrosion control. Ductile iron pipe, in certain environments, is subject to external corrosion.Linear Pola...

  11. Comparison Between Stress Obtained by Numerical Analysis and In-Situ Measurements on a Flexible Pipe Subjected to In-Plane Bending Test

    DEFF Research Database (Denmark)

    Vestergaard Lukassen, Troels; Glejbøl, Kristian; Lyckegaard, Anders

    2016-01-01

    to stress patterns obtained during in-situ OMS measurements carried out during an actual experimental inplane bending test. The study showed a good correlation between the stress variation predicted with the finite element model and the measured stress variation.......To predict the lifetime and long-term properties of tensile armour wires in a dynamically loaded pipe, it is essential to have a tool which allows detailed prediction of the stress variations in the tensile armour wires during global pipe loading. Furthermore, detailed understanding of the stress...... variations will allow for performance optimization of the armour layers. To study the detailed stress variations in flexible pipes during dynamic loading, a comprehensive three-dimensional implicit nonlinear finite element model has been developed. The predicted numerical stress variations will be compared...

  12. LOFT/L3-6, Loss of Fluid Test, 6. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the sixth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were running. The experiment was conducted on 10 December 1980

  13. LOFT/L3-5, Loss of Fluid Test, 5. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1991-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the fifth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were shut off. The experiment was conducted on 29 September 1980

  14. LOFT/L2-5, Loss of Fluid Test, 3. NRC L2 Large Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the third of the NRC L2 Series of nuclear large Break LOCA experiments, conducted on 16 June 1981. It simulated a 100% cold leg break with a maximum heat generation of 40 kW/m and rapid pump coast down

  15. LOFT/L2-3, Loss of Fluid Test, 2. NRC L2 Large Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the second of the NRC L2 Series of nuclear large Break LOCA experiments, and was conducted on 12 May 1979. It simulated a 100% cold leg break with a maximum heat generation of 39 kW/m

  16. 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

  17. 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.

  18. 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.)

  19. 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)

  20. 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.)

  1. Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion with Validation from Mechanical Testing

    International Nuclear Information System (INIS)

    Anderson, Michael T.; Cinson, Anthony D.; Crawford, Susan L.; Doctor, Steven R.; Moran, Traci L.; Watts, Michael W.

    2010-01-01

    Studies at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, are being conducted to evaluate nondestructive examinations (NDE) coupled with mechanical testing of butt fusion joints in high-density polyethylene (HDPE) pipe for assessing lack of fusion. The work provides information to the U.S. Nuclear Regulatory Commission (NRC) on the effectiveness of volumetric inspection techniques of HDPE butt fusion joints in Section III, Division 1, Class 3, buried piping systems in nuclear power plants. This paper describes results from preliminary assessments using ultrasonic and microwave nondestructive techniques and mechanical testing with the high-speed tensile impact test and the side-bend test for determining joint integrity. A series of butt joints were fabricated in 3408, 12-in. IPS DR-11 HDPE material by varying the fusion parameters to create good joints and joints containing a range of lack-of-fusion conditions. Six of these butt joints were volumetrically examined with time-of-flight diffraction (TOFD), phased-array (PA) ultrasound, and the Evisive microwave system. The outer-diameter weld beads were removed for the microwave inspection. In two of the four pipes, both the outer and inner weld beads were removed and the pipe joints re-evaluated. The pipes were sectioned and the joints destructively evaluated with the side-bend test by cutting portions of the fusion joint into slices that were planed and bent. The last step in this limited study will be to correlate the fusion parameters, nondestructive, and destructive evaluation results to validate the effectiveness of what each NDE technology detects and what each does not detect. The results of the correlation will be used in identifying any future work that is needed.

  2. Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

    Science.gov (United States)

    Sabotinov, Luben; Chevrier, Patrick

    The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

  3. 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...

  4. UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

  5. Revisiting LOFT L2-5 large break test in BEMUSE project context. Sensitivity studies

    International Nuclear Information System (INIS)

    Perez, Marina; Batet, Lluis; Pretel, Carme; Reventos, Francesc

    2005-01-01

    Full text of publication follows: Best estimate codes simulate NPPs behavior in principle without any special conservative assumptions. Due to several factors like code solution methods or user effects, the output parameters calculated have an uncertainty associated. The quantification of the these uncertainties becomes crucial when a safety statement is to be made. It is in this scope that GAMA group from CSNI (OECD/NEA) proposed the international BEMUSE project (Best Estimate - Uncertainty and Sensitivity Evaluation) having as main objective the evaluation of different methodologies for the uncertainty and sensitivity analysis of best-estimate code calculations. A number of methodologies prepared in different countries are used in the development of the project activities. The program work consists of 6 phases and currently the first two have already been concluded. Phase II consists in revisiting the ISP-13, the LOFT loss of coolant experiment L2-5 which simulated a double ended 200% cold leg break of a commercial PWR simultaneous with a loss of site power. In order to connect phase II with phase III, in which the uncertainty analysis will be carried out, quite a large number of sensitivity analysis have been performed by simulating system failures and varying fuel elements parameters among others. The presentation will focus on the results of the sensitivity analysis as well as its importance with regards to the uncertainty studies. The methodology used by UPC team was developed by ENUSA and the work is supported by the Spanish regulatory organization. (authors)

  6. Stress Analysis and Model Test of Rock Breaking by Arc Blade Wedged Hob

    Directory of Open Access Journals (Sweden)

    Ying-chao Liu

    2016-07-01

    Full Text Available Based on rock compression-shear damage theory, the mechanical characteristics of an arc blade wedged hob were analyzed to study the rock fragmentation mechanism of hob during excavation, and rock fragmentation forecasting model of the arc blade wedged hob was improved. A spoke type cutter model which is similar to the tunnel boring machine (TBM cutter head was designed to study the rock fragmentation efficiency in different cutter spacing by adjusting the bearing sleeve size to obtain different distances between the hobs. The results show that the hob-breaking rock force mainly comes from three directions. The vertical force along the direction of the tunnel excavation, which is associated with uniaxial compressive strength of rock mass, plays a key role in the process of rock fragmentation. Field project data shows that the prediction model’s results of rock fragmentation in this paper are closer to the measured results than the results of the traditional linear cutting model. The optimal cutter spacing exists among different cutter spacings to get higher rock fragmentation rate and lower energy consumption during rock fragmentation. It is of great reference significance to design the arc blade wedged hob and enhance the efficiency of rock fragmentation in rock strata.

  7. Testing the reliability of supply pipings of the pump-fed power station Markersbach

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, W; Perout, A; Hellwig, K

    1981-01-01

    The leakage tests for the pressurised part of steam raising units is carried out during erection of steam raising units by halogen leak indicators. The effect which is produced by spontaneous change of alkali emission of a glowing anode if it comes into contact with molecules containing halogens is used. The erection leak detection equipment which is independent of the mains supply, works on the principle of high voltage corona discharge. Other equipment is also described briefly. All the equipment described uses dichlorfluormethane (CF/sub 2/Cl/sub 2/) as the halogen test gas - with the commercial name Fridohna 12. The test pressure in the pressurised part of the steam raising unit is 0.5 MPa and the mixing ratio of Fridohna 12 to compressed air is 0.12 kg of Fridohna 12 to 1 cubic metre of atmospheric air. The halogen leak test is superior to the water pressure test.

  8. Development of new damping devices for piping

    International Nuclear Information System (INIS)

    Kobayashi, Hiroe

    1991-01-01

    An increase of the damping ratio is known to be very effective for the seismic design of a piping system. Increasing the damping ratio and reducing the seismic response of the piping system, the following three types of damping devices for piping systems are introduced: (1) visco-elastic damper, (2) elasto-plastic damper and (3) compact dynamic damper. The dynamic characteristics of these damping devices were investigated by the component test and the applicability of them to the piping system was confirmed by the vibration test using a three dimensional piping model. These damping devices are more effective than mechanical snubbers to reduce the vibration of the piping system. (author)

  9. Automated Break-Out Box for use with Low Cost Spacecraft Integration and Test, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Electrical checkout and testing is a critical part of the overall spacecraft integration and test flow. Verifying proper harness and connector signal interfaces is...

  10. 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%.

  11. Breaking Bat

    Science.gov (United States)

    Aguilar, Isaac-Cesar; Kagan, David

    2013-01-01

    The sight of a broken bat in Major League Baseball can produce anything from a humorous dribbler in the infield to a frightening pointed projectile headed for the stands. Bats usually break at the weakest point, typically in the handle. Breaking happens because the wood gets bent beyond the breaking point due to the wave sent down the bat created…

  12. IPM Pipe

    Science.gov (United States)

    Submit A Report View Reports List [+] View Reports Map [+] CDM Alert System Sign Up For Alerts User Login Annual Epidemic Histories Annual Season Summaries Contact Us ipmPIPE User Login Web Administrator Login

  13. Evaluation of thermal displacement behavior of high temperature piping system in power-up test of HTTR. No. 1 results up to 20 MW operation

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio

    2002-03-01

    Temperature of the primary cooling system of the High Temperature Engineering Test Reactor, HTTR, becomes very high because the coolant temperature at the reactor outlet reaches 950degC, and 400degC at inlet of the reactor. Therefore, it is important to confirm the thermal displacement behavior of the high temperature piping system in the primary cooling system from the viewpoint of the structural integrity. Moreover, newly designed 3-dimensional floating support system is adopted to the cooling system, it is meaningful to verify the thermal displacement behavior of the piping system applied the 3-dimensional floating support system. In the power-up test (up to 20 MW operation), thermal displacement behavior of the high temperature piping system was measured. This paper describes the experimental and analytical results of thermal displacement characteristics of the high temperature piping system. The results showed that the resistance force induced from the supporting system effects to the thermal displacement behavior of cooling system, and the analytical results have a good agreement with the experimental results by optimizing the resistant force of the floating support system. Additionally, structural integrity at the 30 MW operation was confirmed by the analysis. (author)

  14. 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.

  15. An evaluation of detection ability of ultrasonic testing with a large aperture transducer for axial cracks in cast stainless steel pipe welds

    International Nuclear Information System (INIS)

    Nishikawa, Yoshito; Ishida, Hitoshi; Kurozumi, Yasuo

    2013-01-01

    Ultrasonic testing is difficult to apply to cast stainless steel which is the material of the main coolant pipes in pressurized water reactors, because of the large attenuation and scattering of ultrasonic waves caused by its macro structure. In this study, ultrasonic testing for progression of axial fatigue cracks of a welded area in the test piece of cast stainless steel pipe was performed using double big-size ultrasonic probes which were formerly developed in INSS. It was found that detection of defects that were over 6% of the target depth for the specimen thickness of 69mm is possible, and detection of defects with over 10% of the target depth is possible for all test conditions. (author)

  16. Water pipe (Shisha, Hookah, Arghile) Smoking and Secondhand Tobacco Smoke Effects on CYP1A2 and CYP2A6 Phenotypes as Measured by Caffeine Urine Test.

    Science.gov (United States)

    Yılmaz, Şenay Görücü; Llerena, Adrián; De Andrés, Fernando; Karakaş, Ümit; Gündoğar, Hasan; Erciyas, Kamile; Kimyon, Sabit; Mete, Alper; Güngör, Kıvanç; Özdemir, Vural

    2017-03-01

    Public policies to stop or reduce cigarette smoking and exposure to secondhand smoke and associated diseases have yielded successful results over the past decade. Yet, the growing worldwide popularity of another form of tobacco consumption, water pipe smoking, has received relatively less attention. To the best of our knowledge, no study to date has evaluated the effects of water pipe smoking on cytochrome P450 (CYP450) activities and drug interaction potential in humans, whereas only limited information is available on the impact of secondhand smoke on drug metabolism. In a sample of 99 healthy volunteers (28 water pipe smokers, 30 secondhand tobacco smoke exposed persons, and 41 controls), we systematically compared CYP1A2 and CYP2A6 enzyme activities in vivo using caffeine urine test. The median self-reported duration of water pipe smoking was 7.5 h/week and 3 years of exposure in total. The secondhand smoke group had a median of 14 h of self-reported weekly exposure to tobacco smoke indoor where a minimum of five cigarettes were smoked/hour for a total of 3.5 years (median). Analysis of variance did not find a significant difference in CYP1A2 and CYP2A6 activities among the three study groups (p > 0.05). Nor was there a significant association between the extent of water pipe or secondhand smoke exposure and the CYP1A2 and CYP2A6 activities (p > 0.05). Further analysis in a subsample with smoke exposure more than the median values also did not reveal a significant difference from the controls. Although we do not rule out an appreciable possible impact of water pipe smoke and secondhand smoke on in vivo activities of these two drug metabolism pathways, variability in smoke constituents from different tobacco consumption methods (e.g., water pipe) might affect drug metabolism in ways that might differ from that of cigarette smoke. Further studies in larger prospective samples are recommended to evaluate water pipe and secondhand tobacco smoke effects

  17. 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

  18. Furnace testing of electrical and pipe-penetration seals based on foamed silicone elastomer: 60, 90, and 120-minute fire ratings

    International Nuclear Information System (INIS)

    Brown, A.

    1979-03-01

    Fire tests of foamed silicone seals for electrical and pipe penetrations have been performed using a furnace with temperature control as heat source. The tests were performed in principle in accordance with the requirements of NORDTEST 5A (ISO 834). The purpose of the tests was to obtain appropriate fire ratings for different seal thicknesses. The report covers. - Description of material used to prepare the seals and method of application - Description of furnace test assembly and method of performing test - Listing of penetrating elements and of the thermocouple array used to measure temperature - Curves of thermocouple readouts and photographs of seals during and after completion of the test. (author)

  19. Aircraft Thermal Management Using Loop Heat Pipes: Experimental Simulation of High Acceleration Environments Using the Centrifuge Table Test Bed (Postprint)

    National Research Council Canada - National Science Library

    Fleming, Andrew J; Leland, Quinn H; Yerkes, Kirk L; Elston, Levi J; Thomas, Scott K

    2006-01-01

    The objective of this paper is to describe the design of an experiment that will examine the effects of elevated acceleration environments on a high-temperature, titanium-water loop heat pipe for actuator cooling...

  20. Thermo-Mechanical Test of Seal System in Flexible Pipe End Fittings

    DEFF Research Database (Denmark)

    Banke, Lars

    1999-01-01

    are driven radially into the barrier layer and supported by the surrounding steel casing. In order to verify the integrity of the concept the seal system is subjected cyclic pressure and temperature variations to simulate the service conditions.The aim of the testing is to demonstrate the sensitivity...... of the seal system geometry and its tolerances necessary to maintain a tight seal. The test is carried out in a purpose built autoclave, in which the seal system can be tested while undergoing variations in pressure and temperature.The paper will present a study on the importance of the geometry of the gasket...... and the inner liner. The inner and outer diameter of the gasket are varied to see the effectiveness of the seal mechanism. The effect of varying the width of the gasket as well as the surface roughness of the components in the seal system is analysed. Finally, it is investigated how the seal system is affected...

  1. 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.

  2. Load tests with a pipe bend DN 425, applying slowly changing bending loads up to occurrence of leak

    International Nuclear Information System (INIS)

    Uhlmann, D.; Hunger, H.

    1990-01-01

    The experimental program deals with the formation of incipient cracks and subsequent crack growth of axially oriented cracks at a pipe bend with a nominal width of DN 425. The pipe bend consists of the ferritic material 20MnMoNi55. The numerical experiments by means of 3 D-FE analyses concentrate on determining the influence of the asymmetric crack depths at the two bend halves, and of the multiple crack fields, on the effective crack strain. (DG) [de

  3. The applicability of the Pennsylvania Notch Test for a new generation of PE pipe grades

    Czech Academy of Sciences Publication Activity Database

    Nezbedová, E.; Hutař, Pavel; Zouhar, Michal; Knésl, Zdeněk; Sadílek, J.; Náhlík, Luboš

    2013-01-01

    Roč. 32, č. 1 (2013), s. 106-114 ISSN 0142-9418 R&D Projects: GA ČR(CZ) GAP108/12/1560; GA MŠk(CZ) ED1.1.00/02.0068 Institutional support: RVO:68081723 Keywords : PENT test * Lifetime prediction * HDPE * Slow crack growth * Fracture mechanics Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 1.816, year: 2013

  4. 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)

  5. First intermediate break test 6IB1 data comparison with a TRAC-BD1/MOD1 blind calculation

    International Nuclear Information System (INIS)

    Wheatley, P.D.

    1985-04-01

    TRAC-BD1/MOD1 has been used to calculate the behavior in the FIST (Full Integral Test Facility) facility during an intermediate break in one of the recirculation loops. Results of the calculation are compared with the data from the experiment, and the analysis is discussed in this report. The calculation was blind with only the initial and boundary conditions available prior to performance of the calculation. The calculation has been previously documented without reference to the experimental data (i.e., prior to release of the data). This report extends the prior report by discussing the analysis of the data to code comparisons. This work was performed as part of the Nuclear Regulatory Commission's support to the FIST program which is being provided at the Idaho National Engineering Laboratory

  6. 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

  7. 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.

  8. RELAP5/MOD2 blind calculation of GERDA small break test and data comparison

    International Nuclear Information System (INIS)

    Ogden, D.M.; Steiner, J.L.; Waterman, M.E.

    1985-01-01

    The Idaho National Engineering Laboratory (INEL), in support of the USNRC, has developed a RELAP5/MOD2 model of the GERDA facility to be used for analysis of the GERDA data, particularly relative to the phenomena of natural circulation and the boiler condenser mode of heat transfer. A blind calculation of GERDA Test 1605AA and a preliminary comparison with experimental data has been performed. The GERDA facility is a single loop integral facility with an electrically heated core. A general arrangement diagram of the facility is shown. The GERDA facility was designed for the performance of both separate effects and overall systems tests

  9. Assessment of SPACE code for multiple failure accident: 1% Cold Leg Break LOCA with HPSI failure at ATLAS Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Hyuk; Lee, Seung Wook; Kim, Kyung-Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Design extension conditions (DECs) is a popular key issue after the Fukushima accident. In a viewpoint of the reinforcement of the defense in depth concept, a high-risk multiple failure accident should be reconsidered. The target scenario of ATLAS A5.1 test was LSTF (Large Scale Test Facility) SB-CL-32 test, a 1% SBLOCA with total failure of high pressure safety injection (HPSI) system of emergency core cooling system (ECCS) and secondary side depressurization as the accident management (AM) action, as a counterpart test. As the needs to prepare the DEC accident because of a multiple failure of the present NPPs are emphasized, the capability of SPACE code, just like other system analysis code, is required to expand the DEC area. The objectives of this study is to validate the capability of SPACE code for a DEC scenario, which represents multiple failure accident like as a SBLOCA with HPSI fail. Therefore, the ATLAS A5.1 test scenario was chosen. As the needs to prepare the DEC accident because of a multiple failure of operating NPPs are emphasized, the capability of SPACE code is needed to expand the DEC area. So the capability of SPACE code was validated for one of a DEC scenario. The target scenario was selected as the ATLAS A5.1 test, which is a 1% SBLOCA with total failure of HPSI system of ECCS and secondary side depressurization. Through the sensitivity study on discharge coefficient of break flow, the best fit of integrated mass was found. Using the coefficient, the ATLAS A5.1 test was analyzed using the SPACE code. The major thermal hydraulic parameters such as the system pressure, temperatures were compared with the test and have a good agreement. Through the simulation, it was concluded that the SPACE code can effectively simulate one of multiple failure accidents like as SBLOCA with HPSI failure accident.

  10. 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.)

  11. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  12. The influence of renewable and non-renewable energy consumption and real income on CO2 emissions in the USA: evidence from structural break tests.

    Science.gov (United States)

    Dogan, Eyup; Ozturk, Ilhan

    2017-04-01

    The objective of this study is to explore the influence of the real income (GDP), renewable energy consumption and non-renewable energy consumption on carbon dioxide (CO 2 ) emissions for the United States of America (USA) in the environmental Kuznets curve (EKC) model for the period 1980-2014. The Zivot-Andrews unit root test with a structural break and the Clemente-Montanes-Reyes unit root test with a structural break report that the analyzed variables become stationary at first-differences. The Gregory-Hansen cointegration test with a structural break and the bounds testing for cointegration in the presence of a structural break show CO 2 emissions, the real income, the quadratic real income, renewable and non-renewable energy consumption are cointegrated. The long-run estimates obtained from the ARDL model indicate that increases in renewable energy consumption mitigate environmental degradation whereas increases in non-renewable energy consumption contribute to CO 2 emissions. In addition, the EKC hypothesis is not valid for the USA. Since we use time-series econometric approaches that account for structural break in the data, findings of this study are robust, reliable and accurate. The US government is advised to put more weights on renewable sources in energy mix, to support and encourage the use and adoption of renewable energy and clean technologies, and to increase the public awareness of renewable energy for lower levels of emissions.

  13. MODEL TESTS AND 3D ELASTIC FINITE ELEMENT ANALYSIS FOR STEEL PIPE PILES WITH WINGS IN STALLED IN SOIL CEMENT COLUMN

    Science.gov (United States)

    Tamai, Toshiyuki; Teramoto, Shuntarou; Kimura, Makoto

    Steel pipe piles with wings installed in soil cement column is a composite foundation of pile consisting of soil improvement with cement and steel pipe with wings. This type of pile shows higher vertical bearing capacity when compared to steel pipe piles that are installed without soil cement. It is thought the wings contribute to higher bearing capacity of this type of piles. The wings are also thought to play the role of structural unification of pile foundations and load transfer. In this study, model test and 3D elastic finite element analysis was carried out in order to elucidate the effect of wings on the structural unification of pile foundation and the load transfer mechanism. Firstly, the model test was carried out in order to grasp the influence of pile with and without wings, the shape of wings of the pile and the unconfined compression strength of the soil cement on the structural unification of the pile foundation. The numerical analysis of the model test was then carried out on the intermediate part of the pile foundation with wings and mathematical model developed. Finally load tran sfer mechanism was checked for the entire length of the pile through this mathematical model and the load sharing ratio of the wings and stress distribution occurring in the soil cement clarified. In addition, the effect of the wing interval on the structural unification of the pile foundation and load transfer was also checked and clarified.

  14. TRAC-PF1/MOD1 independent assessment: Semiscale Mod-2A intermediate break test S-IB-3

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1986-02-01

    The TRAC-PF1/MOD1 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various system codes to predict the detailed thermal/hydraulic response of light water reactors during accident and off-normal conditions. The TRAC code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, an intermediate break test (S-IB-3), performed at the Semiscale Mod-2A facility, has been analyzed. Using an input model with a 3-D VESSEL component, the vessel and downcomer inventories during 3-IB-3 were generally well predicted, but the core heatup was underpredicted compared to data. An equivalent calculation with an all 1-D input model ran about twice as fast as our basecase analysis using a 3-D VESSEL in the input model, but the results of the two calculations diverged significantly for many parameters of interest, with the 3-D VESSEL model results in better agreement with data. 22 refs., 100 figs

  15. LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  16. Discussions on Laboratory Testing of Backwards Erosion Piping of Soil: An Interview with John H. Schmertmann and Frank C. Townsend

    Science.gov (United States)

    2014-11-01

    India: Lahore. Lane, E. W. 1935. Security from under-seepage masonry dams on earth foundations. Transactions of ASCE 100:1235–1272. Richards, K. S...and K. R. Reddy. 2007. Critical appraisal of piping phenomena in earth dams. Bulletin of Engineering Geology and the Environment 66(4):381–402...and I quoted your description in my paper [(Schmertmann 2000)], that it’s a series of little landslides . A kind of flow slide at the end of the pipe

  17. 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.

  18. Breaking Health Insurance Knowledge Barriers Through Games: Pilot Test of Health Care America

    Science.gov (United States)

    James, Juli

    2017-01-01

    Background Having health insurance is associated with a number of beneficial health outcomes. However, previous research suggests that patients tend to avoid health insurance information and often misunderstand or lack knowledge about many health insurance terms. Health insurance knowledge is particularly low among young adults. Objective The purpose of this study was to design and test an interactive newsgame (newsgames are games that apply journalistic principles in their creation, for example, gathering stories to immerse the player in narratives) about health insurance. This game included entry-level information through scenarios and was designed through the collation of national news stories, local personal accounts, and health insurance company information. Methods A total of 72 (N=72) participants completed in-person, individual gaming sessions. Participants completed a survey before and after game play. Results Participants indicated a greater self-reported understanding of how to use health insurance from pre- (mean=3.38, SD=0.98) to postgame play (mean=3.76, SD=0.76); t71=−3.56, P=.001. For all health insurance terms, participants self-reported a greater understanding following game play. Finally, participants provided a greater number of correct definitions for terms after playing the game, (mean=3.91, SD=2.15) than they did before game play (mean=2.59, SD=1.68); t31=−3.61, P=.001. Significant differences from pre- to postgame play differed by health insurance term. Conclusions A game is a practical solution to a difficult health issue—the game can be played anywhere, including on a mobile device, is interactive and will thus engage an apathetic audience, and is cost-efficient in its execution. PMID:29146564

  19. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

    2017-08-15

    An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  20. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2017-08-01

    Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  1. 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

  2. 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.)

  3. Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7

    International Nuclear Information System (INIS)

    Sahota, M.S.

    1982-01-01

    Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

  4. 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

  5. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. TRAC-PF1/MOD 1 independent assessment: Semiscale MOD-2A feedwater-line break (S-SF-3) and steam-line break (S-SF-5) tests

    International Nuclear Information System (INIS)

    Dobranich, D.

    1985-11-01

    The TRAC-PF1/MOD1 independent assessment project at Sandia is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. As part of this effort, calculations for Semiscale Mod-2A test S-SF-3, a feedwater-line break test, and S-SF-5, a steam-line break test, were performed with TRAC-PF1/MOD1. Most aspects of both the S-SF-3 and S-SF-5 steady-state calculations were found to be in good agreement with data. However, the need for a better steam separator model was identified from the S-SF-3 calculation. Overall, the qualitative behavior of both transients was calculated reasonably well; however, there were some discrepancies in the prediction of the quantitative behavior. The results for the S-SF-3 transient calculation indicate that the primary-to-secondary heat transfer degradation began too early. This was possibly due to overprediction of entrainment in the steam generator boiler, leading to an incorrect calculation of the secondary-side fluid distribution during the steady state. However, there was insufficient data to verify this. Results for the S-SF-5 transient calculation indicate that the primary-side fluid temperature response to a steam-line break was in reasonable agreement with data but the pressure response did not coincide with the data. Uncertainties in the data are sufficient to prevent us from determining the exact cause of this discrepancy, but there is indirect evidence that the calculated rate of phase change in the pressurizer was incorrect. 16 refs., 73 figs., 13 tabs

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. Piping vibrations measured during FFTF startup

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1981-03-01

    An extensive vibration survey was conducted on the Fast Flux Test Facility piping during the plant acceptance test program. The purpose was to verify that both mechanical and flow induced vibration amplitudes were of sufficiently low level so that pipe and pipe support integrity would not be compromised over the plant design lifetime. Excitation sources included main heat transport sodium pumps, reciprocating auxiliary system pumps, EM pumps, and flow oscillations. Pipe sizes varied from one-inch to twenty-eight-inches in diameter. This paper describes the test plan; the instrumentation and procedures utilized; and the test results

  13. Effects of high temperature ECC injection on small and large break BWR LOCA simulation tests in ROSA-III program (RUNs 940 and 941)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Kumamaru, Hiroshige; Anoda, Yoshinari; Yonomoto, Taisuke; Murata, Hideo; Tasaka, Kanji

    1990-03-01

    The ROSA-III program, of which principal results are summarized in a report of JAERI 1307, conducted small and large-break loss-of-coolant experiments (RUNs 940 and 941) with high water temperature of the emergency core cooling system (ECCS) are one of the parametric study with respect to the ECCS effect on core cooling. This report presents all the experiment results of these two tests and describes additional finding with respect to the hot ECC effects on core cooling phenomena. By comparing these two tests (water temperature of 393 K) with the standard ECC tests of RUNs 922 and 926 (water temperature of 313 K), it was found that the ECC subcooling variation had a small influence on the core cooling phenomena in 5 % small break tests but had larger influence on them in 200 % break tests. The ECC subcooling effects described in the previous report are reviewed and the temperature distribution in the pressure vessel is investigated for these four tests. (author)

  14. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1982-01-01

    A comprehensive, up-to-date coverage of the theory, design and manufacture of heat pipes and their applications. This latest edition has been thoroughly revised, up-dated and expanded to give an in-depth coverage of the new developments in the field. Significant new material has been added to all the chapters and the applications section has been totally rewritten to ensure that topical and important applications are appropriately emphasised. The bibliography has been considerably enlarged to incorporate much valuable new information. Thus readers of the previous edition, which has established

  15. Determination of hoop direction effective elastic moduli of non-circular profile, fiber reinforced polymer composite sewer liner pipes from lateral ring compression tests

    International Nuclear Information System (INIS)

    Czél, Gergely; Takács, Dénes

    2015-01-01

    A new material property determination method is presented for the calculation of effective elastic moduli of non-circular ring specimens cut from filament wound oval profile polymer composite sewer liner pipes. The hoop direction elastic moduli was determined using the test results obtained from ring compression tests, which is a very basic setup, and requires no special equipment. Calculations were executed for many different oval profiles, and diagrams were constructed, from which the cross section dependent C_e_f_f constants can be taken. The new method was validated by the comparison of tests and finite element analysis results. The calculation method and the diagrams are essential design tools for engineers, and a big step forward in sizing non-circular profile liner pipes. - Highlights: • A simple modulus measurement method is presented for non-circular ring specimens. • The evaluation method is validated against a finite element model. • Profile shape dependent constants are presented for a wide range of cross-sections. • A set of charts with the constants are provided to aid design engineers.

  16. The Break

    DEFF Research Database (Denmark)

    Strand, Anete Mikkala Camille

    2018-01-01

    storytelling to enact fruitful breakings of patterns unbecoming. The claim being, that the hamster wheel of Work-life anno 2016 needs reconfiguration and the simple yet fruitful manner by which this is done is through acknowledging the benefits of bodies, spaces and artifacts – and the benefits of actually...... taking a break, discontinuing for a moment in order to continue better, wiser and more at ease. Both within and as part of the daily routines, and – now and then – outside these routines in the majesty of nature with time to explore and redirect the course of life in companionships with fellow man...

  17. Minimal duality breaking in the Kallen-Lehman approach to 3D Ising model: A numerical test

    International Nuclear Information System (INIS)

    Astorino, Marco; Canfora, Fabrizio; Martinez, Cristian; Parisi, Luca

    2008-01-01

    A Kallen-Lehman approach to 3D Ising model is analyzed numerically both at low and high temperatures. It is shown that, even assuming a minimal duality breaking, one can fix three parameters of the model to get a very good agreement with the Monte Carlo results at high temperatures. With the same parameters the agreement is satisfactory both at low and near critical temperatures. How to improve the agreement with Monte Carlo results by introducing a more general duality breaking is shortly discussed

  18. Simulation of phenomena at crack-like leaks and breaks in piping with consideration of fluid-structure interaction. Final report; Simulation der Phaenomene bei rissartigen Lecks und Bruechen in Rohrleitungen unter Beruecksichtigung der Fluid-Struktur-Kopplung. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Sievers, J.; Grebner, H.; Bahr, L.; Heckmann, K.; Arndt, J.; Pallas-Moner, G.

    2013-11-15

    The evaluation of fluid flow rates through crack-like leaks in pressurized components plays an important role for assessments on break preclusion, especially leak-before-break considerations. In the framework of project RS1194 various calculation methods for the simulation of structure mechanical and thermo-hydraulic phenomena due to flows through crack-like leaks in the coolant circuit were examined and validated on selected leak rate experiments. Besides large program systems as ATHLET, CFX and ADINA also several simplified evaluation methods included in the GRS program WinLeck were applied especially for the determination of leak rates. For the validation of analysing methods, tests were selected previously conducted at the former Nuclear Research Centre (KfK) at Karlsruhe and the Power Plant Union (KWU). The review of experimental results already at disposal in regards to availability of measure d values of thermo-hydraulic parameters like flow-through rates, spatial distributions of pressure, temperature and aggregate state of the medium, velocity of the medium as well as leak openings, displacements and structure strains indicated, that the experiments in terms of quantification of thermo-hydraulic and structure mechanical phenomena as well as appropriate coupling effects do not provide sufficiently meaningful results. Due to missing experiments for validation of 3d numerical flow simulation in crack-like leaks experiments with flow through a Venturi orifice, which are relevant in this context, were chosen. Experiments with single phase flow were considered as well as ones with two phase flow. The post-calculations of the single phase flow showed a good agreement between the calculation results and the appropriate measured data. In the two phase flow, despite tests with various model variations, no satisfying agreement between calculation and test could be reached. According to the authors' opinion is the model approach available in CFX for the

  19. Investigation on vibrational evaluation criteria for small-bore pipe

    International Nuclear Information System (INIS)

    Tsuji, Takashi; Maekawa, Akira; Takahashi, Tsuneo; Kato, Minoru; Torigoe, Yuichi

    2013-01-01

    The well-known organization such as API and SwRI in USA developed criteria for piping vibrational evaluation. These criteria are targeted for main pipes, but not branch pipes with small bore. In this study, applicability of criteria of API and SwRI to branch pipes was investigated. Vibration test using piping system with small bore branch pipe was conducted and amplitudes of vibrational stress and displacement were measured for various exciting force. In comparison of the measurements with the two criteria, though the criteria of API and SwRI were applicable to small bore branch pipe, they made too conservative evaluation. (author)

  20. 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

  1. 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

  2. Theoretical investigations into the influence of the position of a breaking line on the tensile failure of flat, round, bevel-edged tablets using finite element methodology (FEM) and its practical relevance for industrial tablet strength testing.

    Science.gov (United States)

    Podczeck, Fridrun; Newton, J Michael; Fromme, Paul

    2014-12-30

    Flat, round tablets may have a breaking ("score") line. Pharmacopoeial tablet breaking load tests are diametral in their design, and industrially used breaking load testers often have automatic tablet feeding systems, which position the tablets between the loading platens of the machine with the breaking lines in random orientation to the applied load. The aim of this work was to ascertain the influence of the position of the breaking line in a diametral compression test using finite element methodology (FEM) and to compare the theoretical results with practical findings using commercially produced bevel-edged, scored tablets. Breaking line test positions at an angle of 0°, 22.5°, 45°, 67.5° and 90° relative to the loading plane were studied. FEM results obtained for fully elastic and elasto-plastic tablets were fairly similar, but they highlighted large differences in stress distributions depending on the position of the breaking line. The stress values at failure were predicted to be similar for tablets tested at an angle of 45° or above, whereas at lower test angles the predicted breaking loads were up to three times larger. The stress distributions suggested that not all breaking line angles would result in clean tensile failure. Practical results, however, did not confirm the differences in the predicted breaking loads, but they confirmed differences in the way tablets broke. The results suggest that it is not advisable to convert breaking loads obtained on scored tablets into tablet tensile strength values, and comparisons between different tablets or batches should carefully consider the orientation of the breaking line with respect to the loading plane, as the failure mechanisms appear to vary. Copyright © 2014 Elsevier B.V. All rights reserved.

  3. The experimental test of the adequateness of relativistic impulse approximation when describing the lightest nuclei break-up

    International Nuclear Information System (INIS)

    Sitnik, I.M.

    1995-01-01

    The behaviour of the lightest nuclei break-up cross sections at zero angle has been analyzed in vicinity of the maximum. It is shown that asymmetry of cross sections relatively maximum is in conflict with nonrelativistic impulse approximation, but agrees well with one of relativistic approaches to describe this process. 10 refs., 9 figs

  4. 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

  5. Simulation of an SBLOCA Test of Shutdown Cooling System Line Break with the SMARTITL Facility using the MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yeon-Sik; Suh, Jae-Seung [System Engineering and Technology, Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sung-Uk; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    An LBLOCA (Large-Break Loss of Coolant Accident) was inherently eliminated in the design stage. The SMART design has a thermal power of 330MW. Its core exit temperature and pressurizer pressure are 323 .deg. C and 15MPa during normal operating conditions, respectively. An integral-effect test loop for SMART (SMARTITL), called FESTA (Facility for Experimental Simulation of Transients and Accidents), was designed to simulate the integral thermal-hydraulic behavior of SMART. The objectives of SMART-ITL are to investigate and understand the integral performance of reactor systems and components, and the thermal-hydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. SMART-ITL with four steam generators and PRHRS, has an advantage for a multi-loop effect compared with VISTA-ITL with a single loop. The integral-effect test data will also be used to validate the related thermal-hydraulic models of the safety analysis code such as TASS/SMR-S which is used for a performance and accident analysis of the SMART design. In addition, a scoping analysis on the scaling difference between the standard design of SMART and the basic design of SMART-ITL was performed for an SBLOCA (Small-Break Loss of Coolant Accident) scenario using a best-estimate safety analysis code, MARS-KS. This paper introduces a comparison of an SBLOCA test of a shutdown cooling system line break using SMART-ITL with its post-test calculation using the MARS-KS code. An SBLOCA test and its post-test calculation were successfully performed using the SMART-ITL facility and MARS-KS code. The SBLOCA break is a guillotine break, and its location is on the SCS line (nozzle part of the RCP suction). The steady-state conditions were achieved to satisfy the initial test conditions presented in the test requirement and its boundary conditions were properly simulated.

  6. 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)

  7. 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

  8. 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.

  9. Carbon dioxide emissions, economic growth, energy use, and urbanization in Saudi Arabia: evidence from the ARDL approach and impulse saturation break tests.

    Science.gov (United States)

    Raggad, Bechir

    2018-05-01

    This study investigates the existence of long-run relationship between CO 2 emissions, economic growth, energy use, and urbanization in Saudi Arabia over the period 1971-2014. The autoregressive distributed lag (ARDL) approach with structural breaks, where structural breaks are identified with the recently impulse saturation break tests, is applied to conduct the analysis. The bounds test result supports the existence of long-run relationship among the variables. The existence of environmental Kuznets curve (EKC) hypothesis has also been tested. The results reveal the non-validity of the EKC hypothesis for Saudi Arabia as the relationship between GDP and pollution is positive in both the short and the long run. Moreover, energy use increases pollution both in short and long run in the country. On the contrary, the results show a negative and significant impact of urbanization on carbon emissions in Saudi Arabia, which means that urban development is not an obstacle to the improvement of environmental quality. Consequently, policy-makers in Saudi Arabia should consider the efficiency enhancement, frugality in energy consumption, and especially increase the share of renewable energies in the total energy mix.

  10. Diffusion in flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brogaard Kristensen, S.

    2000-06-01

    This report describes the work done on modelling and simulation of the complex diffusion of gas through the wall of a flexible pipe. The diffusion and thus the pressure in annulus depends strongly on the diffusion and solubility parameters of the gas-polymer system and on the degree of blocking of the outer surface of the inner liner due to pressure reinforcements. The report evaluates the basis modelling required to describe the complex geometries and flow patterns. Qualitatively results of temperature and concentration profiles are shown in the report. For the program to serve any modelling purpose in 'real life' the results need to be validated and possibly the model needs corrections. Hopefully, a full-scale test of a flexible pipe will provide the required temperatures and pressures in annulus to validate the models. (EHS)

  11. Diffusion in flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brogaard Kristensen, S

    2000-06-01

    This report describes the work done on modelling and simulation of the complex diffusion of gas through the wall of a flexible pipe. The diffusion and thus the pressure in annulus depends strongly on the diffusion and solubility parameters of the gas-polymer system and on the degree of blocking of the outer surface of the inner liner due to pressure reinforcements. The report evaluates the basis modelling required to describe the complex geometries and flow patterns. Qualitatively results of temperature and concentration profiles are shown in the report. For the program to serve any modelling purpose in 'real life' the results need to be validated and possibly the model needs corrections. Hopefully, a full-scale test of a flexible pipe will provide the required temperatures and pressures in annulus to validate the models. (EHS)

  12. 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

  13. SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

    Energy Technology Data Exchange (ETDEWEB)

    Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

    1995-09-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

  14. Flow induced vibrations of piping

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.

    1977-01-01

    In order to design the supports of piping systems, estimations of the vibrations induced by the fluid conveyed through the pipes are generally needed. For that purpose it is necessary to calculate the model parameters of liquid containing pipes. In most computer codes, fluid effects are accounted for just by adding the fluid mass to the structure. This may lead to serious errors. This paper presents a method to take into account these effects, by solving a coupled mechanical-acoustical problem: the computer code TEDEL of the C.E.A /D.E.M.T. System, based on the finite-elements method, has been extended to calculate simultaneously the pressure fluctuations in the fluid and the vibrations of the pipe. By this way the mechanical-acoustical coupled eigenmodes of any piping system can be obtained. These eigenmodes are used to determine the response of the system to various sources. Equations have been written in the hypohesis that acoustical wave lengths remain large compared to the diameter of the pipe. The method has been checked by an experiment performed on the GASCOGNE loop at D.E.M.T. The piping system under test consists of a tube with four elbows. The circuit is ended at each extremity by a large vessel which performs acoustical isolation by generating modes for the pressure. Excitation of the circuit is caused by a valve located near the downstream vessel. This provides an efficient localised broad band acoustical source. The comparison between the test results and the calculations has shown that the low frequency resonant characteristics of the pipe and the vibrational amplitude at various flow-rates can be correctly predicted

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  16. Low Cost High Performance Generator Technology Program. Volume 5. Heat pipe topical, appendices

    International Nuclear Information System (INIS)

    1975-07-01

    Work performed by Dynatherm Corporation for Teledyne Isotopes during a program entitled ''Heat Pipe Fabrication, Associated Technical Support and Reporting'' is reported. The program was initiated on November 29, 1972; the main objectives were accomplished with the delivery of the heat pipes for the HPG. Life testing of selected heat pipe specimens is continuing to and beyond the present date. The program consisted of the following tasks: Heat Pipe Development of Process Definition; Prototype Heat Pipes for Fin Segment Test; HPG Heat Pipe Fabrication and Testing; Controlled Heat Pipe Life Test; and Heat Pipe Film Coefficient Determination

  17. Mechanized ultrasonic inspection of austenitic pipe systems

    International Nuclear Information System (INIS)

    Dressler, K.; Luecking, J.; Medenbach, S.

    1999-01-01

    The contribution explains the system of standard testing methods elaborated by ABB ZAQ GmbH for inspection of austenitic plant components. The inspection tasks explained in greater detail are basic materials testing (straight pipes, bends, and pipe specials), and inspection of welds and dissimilar welds. The techniques discussed in detail are those for detection and sizing of defects. (orig./CB) [de

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

    International Nuclear Information System (INIS)

    Huang, S.N.

    1981-03-01

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

  19. Development of bore tools for pipe inspection

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Nakahira, Masataka; Taguchi, Kou; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Reactor (ITER), replacement and maintenance on in-vessel components requires that all cooling pipes connected be cut and removed, that a new component be installed, and that all cooling pipes be rewelded. After welding is completed, welded area must be inspected for soundness. These tasks require a new work concept for securing shielded area and access from narrow ports. Tools had to be developed for nondestructive inspection and leak testing to evaluate pipe welding soundness by accessing areas from inside pipes using autonomous locomotion welding and cutting tools. A system was proposed for nondestructive inspection of branch pipes and the main pipe after passing through pipe curves, the same as for welding and cutting tool development. Nondestructive inspection and leak testing sensors were developed and the basic parameters were obtained. In addition, the inspection systems which can move inside pipes and conduct the nondestructive inspection and the leak testing were developed. In this paper, an introduction will be given to the current situation concerning the development of nondestructive inspection and leak testing machines for the branch pipes. (author)

  20. 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

  1. High temperature heat pipe experiments in low earth orbit

    International Nuclear Information System (INIS)

    Woloshun, K.; Merrigan, M.A.; Sena, J.T.; Critchley, E.

    1993-01-01

    Although high temperature, liquid metal heat pipe radiators have become a standard component on most high power space power system designs, there is no experimental data on the operation of these heat pipes in a zero gravity or micro-gravity environment. Experiments to benchmark the transient and steady state performance of prototypical heat pipe space radiator elements are in preparation for testing in low earth orbit. It is anticipated that these heat pipes will be tested aborad the Space Shuttle in 1995. Three heat pipes will be tested in a cargo bay Get Away Special (GAS) canister. The heat pipes are SST/potassium, each with a different wick structure; homogeneous, arterial, and annular gap, the heat pipes have been designed, fabricated, and ground tested. In this paper, the heat pipe designs are specified, and transient and steady-state ground test data are presented

  2. Experimental analytical study on heat pipes

    International Nuclear Information System (INIS)

    Ismail, K.A.R.; Liu, C.Y.; Murcia, N.

    1981-01-01

    An analytical model is developed for optimizing the thickness distribution of the porous material in heat pipes. The method was used to calculate, design and construct heat pipes with internal geometrical changes. Ordinary pipes are also constructed and tested together with the modified ones. The results showed that modified tubes are superior in performance and that the analytical model can predict their performance to within 1.5% precision. (Author) [pt

  3. Ice plugging of pipes using liquid nitrogen

    International Nuclear Information System (INIS)

    Twigg, R.J.

    1987-03-01

    This report presents a study on the ice plugging of pipe using liquid nitrogen, and is based on a literature review and on discussions with individuals who use the technique. Emphasis is placed on ferritic alloys, primarily carbon steels, in pipe sized up to 60 cm in diameter and on austenitic stainless steels in pipe sizes up to 30 cm in diameter. This technique is frequently used for leak testing in nuclear facilities

  4. International crude oil prices and the stock prices of clean energy and technology companies: Evidence from non-linear cointegration tests with unknown structural breaks

    International Nuclear Information System (INIS)

    Bondia, Ripsy; Ghosh, Sajal; Kanjilal, Kakali

    2016-01-01

    Increasing greenhouse gas emissions, exhaustibility and geo-politics induced price volatility of crude oil has magnified the importance of looking for alternative sources of energy. In this paper, we investigate the long term relationship of stock prices of alternative energy companies with oil prices in a multivariate framework. To this end, we use threshold cointegration tests, which endogenously incorporate possible regime shifts in long run relationship of underlying variables. In contrast to the findings of the previous study by Managi and Okimoto (2013), our results indicate presence of cointegration among the variables with two endogenous structural breaks. This study confirms that ignoring the presence of structural breaks in a long time series data, as has been done in previous study, can produce misleading results. In terms of causality, while the stock prices of alternative energy companies are impacted by technology stock prices, oil prices and interest rates in the short run, there is no causality running towards prices of alternative energy stock prices in the long run. The study discusses the possible reasons behind the empirical findings and concludes with a discussion on short run and long run investment opportunities for the investors. - Highlights: • Cointegration between alternative energy companies stock price and oil price. • Threshold cointegration tests are employed. • Cointegration among the variables exists with two endogenous structural breaks. • Alternative energy companies stock price impacted by oil prices in short run. • No causality running towards prices of alternative energy stock prices in long run.

  5. Flow conditions of fresh mortar and concrete in different pipes

    International Nuclear Information System (INIS)

    Jacobsen, Stefan; Haugan, Lars; Hammer, Tor Arne; Kalogiannidis, Evangelos

    2009-01-01

    The variation in fresh concrete flow rate over the pipe cross section was investigated on differently coloured and highly flowable concrete mixes flowing through pipes of different materials (rubber, steel, acryl). First, uncoloured (gray) concrete was poured through the pipe and the pipe blocked. Similar but coloured (black) concrete was then poured into the pipe filled with gray concrete, flowing after the gray concrete for a while before being blocked and hardened. The advance of the colouring along the pipe wall (showing boundary flow rate) was observed on the moulded concrete surface appearing after removing the pipe from the hardened concrete. The shapes of the interfaces between uncoloured and coloured concrete (showing variation of flow rate over the pipe cross section) were observed on sawn surfaces of concrete half cylinders cut along the length axes of the concrete-filled pipe. Flow profiles over the pipe cross section were clearly seen with maximum flow rates near the centre of the pipe and low flow rate at the pipe wall (typically rubber pipe with reference concrete without silica fume and/or stabilizers). More plug-shaped profiles, with long slip layers and less variation of flow rate over the cross section, were also seen (typically in smooth acrylic pipes). Flow rate, amount of concrete sticking to the wall after flow and SEM-images of pipe surface roughness were observed, illustrating the problem of testing full scale pumping.

  6. The analysis of 14.8 percent cold leg break without the application of hydroaccumulators in the PMK-NHV test facility

    International Nuclear Information System (INIS)

    Szabados, L.; Ezsoel, Gy.; Perneczky, L.

    1990-12-01

    A series of reactor safety tests have been performed in the experimental reactor simulation facility PMK-NHV of the Paks Nuclear Power Plant, Hungary, with and without the use of hydroaccumulator (SIT) operation. 14.8 percent cold leg break simulation experiments are reported without SITs in action, and the measurement results were analyzed using the RELAP5/mod2 computer code. The description of the experiment is followed by the parameter variations and their analysis, together with an interpretation of the measurement results. The lessons from the LOCA simulation tests are evaluated. (R.P.) 10 refs.; 48 figs.; 3 tabs

  7. Breaking away.

    Science.gov (United States)

    Innes, G M; Sosnow, P L

    1995-05-01

    While life as hospital employees was comfortable, the lure of independence won out for these two emergency department physicians. Breaking away to develop a new company was not easy, but it's paid off for the entrepreneurs of the Capital Region Emergency Medicine, P.C. Developing an emergency medicine business meant learning all aspects of business: billing services, evaluating legal services, raising capital, and becoming employers. The advantage has been an ability to use profits to improve the moral of staff, an increase in salary, and an overall sense of satisfaction.

  8. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  9. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  10. Strong Electroweak Symmetry Breaking

    CERN Document Server

    Grinstein, Benjamin

    2011-01-01

    Models of spontaneous breaking of electroweak symmetry by a strong interaction do not have fine tuning/hierarchy problem. They are conceptually elegant and use the only mechanism of spontaneous breaking of a gauge symmetry that is known to occur in nature. The simplest model, minimal technicolor with extended technicolor interactions, is appealing because one can calculate by scaling up from QCD. But it is ruled out on many counts: inappropriately low quark and lepton masses (or excessive FCNC), bad electroweak data fits, light scalar and vector states, etc. However, nature may not choose the minimal model and then we are stuck: except possibly through lattice simulations, we are unable to compute and test the models. In the LHC era it therefore makes sense to abandon specific models (of strong EW breaking) and concentrate on generic features that may indicate discovery. The Technicolor Straw Man is not a model but a parametrized search strategy inspired by a remarkable generic feature of walking technicolor,...

  11. 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.

  12. Thermal expansion movements of piping during FFTF plant startup

    International Nuclear Information System (INIS)

    Lindquist, M.R.

    1981-03-01

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

  13. 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...

  14. 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

  15. Detecting pipe bursts by monitoring water demand

    NARCIS (Netherlands)

    Bakker, M.; Vreeburg, J.H.G.; Van der Roer, M.; Sperber, V.

    2012-01-01

    An algorithm which compares measured and predicted water demands to detect pipe bursts was developed and tested on three data sets of water demand and reported pipe bursts of three years. The algorithm proved to be able to detect bursts where the water loss exceeds 30% of the average water demand in

  16. Fatigue analysis of aluminum drill pipes

    Directory of Open Access Journals (Sweden)

    João Carlos Ribeiro Plácido

    2005-12-01

    Full Text Available An experimental program was performed to investigate the fundamental fatigue mechanisms of aluminum drill pipes. Initially, the fatigue properties were determined through small-scale tests performed in an optic-mechanical fatigue apparatus. Additionally, full-scale fatigue tests were carried out with three aluminum drill pipe specimens under combined loading of cyclic bending and constant axial tension. Finally, a finite element model was developed to simulate the stress field along the aluminum drill pipe during the fatigue tests and to estimate the stress concentration factors inside the tool joints. By this way, it was possible to estimate the stress values in regions not monitored during the fatigue tests.

  17. 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.)

  18. Assessment of predictive capability of REFLA/TRAC code for large break LOCA transient in PWR using LOFT L2-5 test data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

    1994-03-01

    The REFLA/TRAC code is a best estimate code developed at Japan Atomic Energy Research Institute (JAERI) to provide advanced predictions of thermal hydraulic transient in light water reactors (LWRs). The REFLA/TRAC code uses the TRAC-PF1/MOD1 code as the framework of the code. The REFLA/TRAC code is expected to be used for the calibration of licensing codes, accident analysis, accident simulation of LWRs, and design of advanced LWRs. Several models have been implemented to the TRAC-PF1/MOD1 code at JAERI including reflood model, condensation model, interfacial and wall friction models, etc. These models have been verified using data from various separate effect tests. This report describes an assessment result of the REFLA/TRAC code, which was performed to assess the predictive capability for integral system behavior under large break loss of coolant accident (LBLOCA) using data from the LOFT L2-5 test. The assessment calculation confirmed that the REFLA/TRAC code can predict break mass flow rate, emergency core cooling water bypass and clad temperature excellently in the LOFT L2-5 test. The CPU time of the REFLA/TRAC code was about 1/3 of the TRAC-PF1/MOD1 code. The REFLA/TRAC code can perform stable and fast simulation of thermal hydraulic behavior in PWR LBLOCA with enough accuracy for practical use. (author)

  19. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  20. Experimental benchmark for piping system dynamic response analyses

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  1. Experimental benchmark for piping system dynamic-response analyses

    International Nuclear Information System (INIS)

    1981-01-01

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

  2. Multiple blowdown pipe experiments with the PPOOLEX facility

    International Nuclear Information System (INIS)

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

    2011-03-01

    This report summarizes the results of the experiments with two steel 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 the blowdown pipes to the condensation pool. The main purpose of the experiment series was to study chugging phenomena (rapid condensation) while steam is discharged through two parallel blowdown pipes into the condensation pool filled with sub-cooled water. Particularly, the aim was to study if the pipe material (polycarbonate) used in the earlier experiment series with two blowdown pipes has had an effect on the general chugging behaviour and measured loads. In the experiments the initial temperature of the pool water was 20 deg. C. The steam flow rate ranged from 220 g/s to 2 350 g/s and the temperature of incoming steam from 148 deg. C to 207 deg. C. The formation and collapse of steam bubbles and the movement of the steam/water interface inside the pipes was non-synchronous. There could be even a 70 ms time difference between the occurrences of steam bubble collapses at the outlets of the two pipes. There was no clear pattern in which pipe the steam bubble first starts to collapse. Several successive bubbles could collapse first in either pipe but then the order changed for a single or several cycles. High pressure loads were measured inside the blowdown pipes due to rapid condensation of the steam volumes in the pipes and resulting water hammer effects. The loads seemed to be higher in pipe 1 than in pipe 2. An explanation for this could be a possible unequal distribution of steam flow between the two pipes. The pipe material has an effect on the condensation phenomena inside the blowdown pipes. A huge difference in the measured pressure curves inside the pipes could be observed compared to the experiments with the polycarbonate pipes. With the same test conditions the amplitude of the

  3. Multiple blowdown pipe experiments with the PPOOLEX facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-15

    This report summarizes the results of the experiments with two steel 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 the blowdown pipes to the condensation pool. The main purpose of the experiment series was to study chugging phenomena (rapid condensation) while steam is discharged through two parallel blowdown pipes into the condensation pool filled with sub-cooled water. Particularly, the aim was to study if the pipe material (polycarbonate) used in the earlier experiment series with two blowdown pipes has had an effect on the general chugging behaviour and measured loads. In the experiments the initial temperature of the pool water was 20 deg. C. The steam flow rate ranged from 220 g/s to 2 350 g/s and the temperature of incoming steam from 148 deg. C to 207 deg. C. The formation and collapse of steam bubbles and the movement of the steam/water interface inside the pipes was non-synchronous. There could be even a 70 ms time difference between the occurrences of steam bubble collapses at the outlets of the two pipes. There was no clear pattern in which pipe the steam bubble first starts to collapse. Several successive bubbles could collapse first in either pipe but then the order changed for a single or several cycles. High pressure loads were measured inside the blowdown pipes due to rapid condensation of the steam volumes in the pipes and resulting water hammer effects. The loads seemed to be higher in pipe 1 than in pipe 2. An explanation for this could be a possible unequal distribution of steam flow between the two pipes. The pipe material has an effect on the condensation phenomena inside the blowdown pipes. A huge difference in the measured pressure curves inside the pipes could be observed compared to the experiments with the polycarbonate pipes. With the same test conditions the amplitude of the

  4. Pipe supports and anchors - LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1983-06-01

    Pipe design and support design can not be treated as separate disciplines. A coordinated design approach is required if LMFBR pipe system adequacy is to be achieved at a reasonable cost. It is particularly important that system designers understand and consider those factors which influence support train flexibility and thus the pipe system dynamic stress levels. The system approach must not stop with the design phase but should continue thru the erection and acceptance test procedures. The factors that should be considered in the design of LMFBR pipe supports and anchors are described. The various pipe support train elements are described together with guidance on analysis, design and application aspects. Post erection acceptance and verification test procedures are then discussed

  5. 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.

  6. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

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

  7. Preventive testing and leakage detection in pipe-lines of steam condensers and generators of a PWR type reactor

    International Nuclear Information System (INIS)

    Canalini, A.; Carvalho, N.C. de

    1985-01-01

    The non-destructive methods: Spum, Helium and Hydrostatic used in leakage detection in condenser pipelines for PWR type reactors are presented. The time, costs, sensitivity, resources necessary and personnel development factors are considered to choose adequated method, in function of nuclear power plant conditions. The leakage tests are applied in pressurized systems or vacuum. Eddy Current testing is used in condensers and steam generators aiming to avoid leakage in these equipments. The spume testing for leakage detection in condenser pipelines - which operation - and hydrostatic testing for leakage detection through reaming with shutdown - were most efficients. The Helium testing applied in pressurized systems or submitted to vacuum systems presented satisfactory results. The Eddy Current testing in condenser and steam generator pipelines reached desired objective, reducing leakage in the first and preserving the integrity in the second. (M.C.K.) [pt

  8. The Break

    DEFF Research Database (Denmark)

    Strand, Anete Mikkala Camille; Larsen, Jens

    2015-01-01

    the challenges of the million-dollar question is stemming from the ‘bets on the future’ – or what David Boje coins as ‘antenarratives’, (Boje, 2008) that emerged through various reconfiguring story actions, on two different occasions. The paper thus elaborates on two cases of restorying events; One taking place...... that language and the social has been granted too much power on the dispense of the bodily, physical and biological – or in short, in dispense of the material. The break To be or not to be poses the theoretical notion of dis-/continuity (Barad, 2007, 2010) from the quantum approach to storytelling (Strand 2012...... in their use of the communicative platform of Object theatre from the methodology of Material Storytelling (Strand 2012). The Bets on the Future piece discusses the extend to which the cases of using this kind of technologies may provide fruitful ‘bets on the future’ in regard to the million-dollar question...

  9. 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.

  10. 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.

  11. 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)

  12. Experimental Investigation on Corrosion of Cast Iron Pipes

    Directory of Open Access Journals (Sweden)

    H. Mohebbi

    2011-01-01

    Full Text Available It is well known that corrosion is the predominant mechanism for the deterioration of cast iron pipes, leading to the reduction of pipe capacity and ultimate collapse of the pipes. In order to assess the remaining service life of corroded cast iron pipes, it is imperative to understand the mechanisms of corrosion over a long term and to develop models for pipe deterioration. Although many studies have been carried out to determine the corrosion behavior of cast iron, little research has been undertaken to understand how cast iron pipes behave over a longer time scale than hours, days, or weeks. The present paper intends to fill the gap regarding the long-term corrosion behaviour of cast iron pipes in the absence of historical data. In this paper, a comprehensive experimental program is presented in which the corrosion behaviour of three exservice pipes was thoroughly examined in three simulated service environments. It has been found in the paper that localised corrosion is the primary form of corrosion of cast iron water pipes. It has also been found that the microstructure of cast irons is a key factor that affects the corrosion behaviour of cast iron pipes. The paper concludes that long-term tests on corrosion behaviour of cast iron pipes can help develop models for corrosion-induced deterioration of the pipes for use in predicting the remaining service life of the pipes.

  13. On Line Enrichment Monitor (OLEM) UF6 Tests for 1.5" Sch40 SS Pipe, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, José A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Garner, Jim [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Younkin, Jim [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Simmons, Darrell W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-01-01

    gas within the unit header pipe as a function of time. The OLEM components have been tested on ORNL UF6 flow loop. Data were collected at five different enrichment levels (0.71%, 2.97%, 4.62%, 6.0%, and 93.7%) at several pressure conditions. The test data were collected in the standard OLEM N.4242 file format for each of the conditions with a 10-minute sampling period and then averaged over the span of constant pressures. Analysis of the collected data has provided enrichment constants that can be used for 1.5” stainless steel schedule 40 pipe measurement sites. The enrichment constant is consistent among all the wide range of enrichment levels and pressures used.

  14. A new multiple channel data recording system for mechanised ultrasonic testing of pipes and nozzles by A-scan processing

    International Nuclear Information System (INIS)

    Heumueller, R.; Rathgeb, W.; Szafarska, E.; Bertus, N.; Erhard, A.; Montag, H.J.; Wuestenberg, H.

    1989-01-01

    A system of equipment for ultrasonic testing in nuclear technique is introduced. This is a four channel ultrasonic equipment, which consists of a manipulator suitable for components, up to four conventional test heads, a test head connection box connected with them via 20 metres of coaxial cable, a documentation unit for signal detection and conversion, a data collection computer for parametricising the equipment, measurement display and representation and a disc memory. The advantages of this test system lie in its easy use because of the compact equipment dimensions, in the data collection of the complete A picture by the documentation unit and in the flexible evaluation of the collected data by the computer. (MM) [de

  15. PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History

    International Nuclear Information System (INIS)

    1987-01-01

    1 - Description of test facility: PHEBUS test facility operated at CEA Research Center Cadarache consists of a pressurized circuit involving pumps, heat exchangers and a blowdown tank - 25 nuclear fuel rod bundle, coupled to a separate driver core; - active length 0.8 m, cosine axial power profile; - pressurized and un-pressurized fuel rods; - controlled cooling conditions at the bundle inlet (blowdown, refill and reflood period); - de-pressurized test rig volume 0.22 m 3 . The following 'as measured' boundary conditions (B.C.) were offered to participants as options with decreasing challenge to their analytical approach: Boundary conditions B.C.0: - full thermal-hydraulic analysis of PHEBUS test rig (was not recommended). Boundary conditions B.C.1: - thermal power level of fuel bundle; - fluid inlet conditions to bundle section. Boundary conditions B.C.2: - local cladding temperatures of rods; - heat transfer coefficients. Boundary conditions B.C.3: - cladding temperatures of rods; - internal pressure of rods. 2 - Description of test: Post-test investigation into the response of a nuclear fuel bundle to a large break loss of coolant accident with respect to - local fuel temperatures, - cladding strain at the time of burst, - time to burst and under given thermal-hydraulic boundary conditions of PHEBUS-test 218

  16. 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

  17. 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.

  18. Testing of in situ and ex situ bioremediation approaches for an oil-contaminated peat bog following a pipeline break

    International Nuclear Information System (INIS)

    Wilson, J.J.; Lee, D.W.; Yeske, B.M.; Kuipers, F.

    2000-01-01

    The feasibility of treating a 1985 pipeline spill of light Pembina Cardium crude oil at a bog near Violet Grove, Alberta was discussed. Pembina Pipeline Corporation arranged for a treatability test to be conducted on oil-contaminated sphagnum peat moss from the site to determine effective in situ or ex situ remediation options for the site. The test was used to evaluate the biodegradation potential of contaminants. Four tests were designed to simulate field different field treatment approaches and to collect critical data on toxicity and leachability of the peat moss. The tests included a bioslurry test, a soil microcosm test, an aerated water saturated peat column test, and a standard toxicity characteristic leachate potential test. The first three tests gave similar results of at least 74 per cent biodegradation of the residual crude oil on the peat solids and no residual toxicity as measured by the Microtox Assay. It was determined that both in situ bioremediation using an aerated water injection system or an ex situ landfarming approach would achieve required criteria and no fertilizers would be necessary to maintain active bioremediation. The new gas-liquid reactor (GLR) aeration technology used in these tests creates a constant supply of hyperoxygenated water prior to column injection. The continuous release of tiny air bubbles maximizes air surface area and increases the gas transfer rates. 3 tabs., 3 figs

  19. Piping system damping data at higher frequencies

    International Nuclear Information System (INIS)

    Ware, A.G.

    1987-01-01

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

  20. An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-03-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

  1. Solar heat-pipe wick modeling

    Energy Technology Data Exchange (ETDEWEB)

    Andraka, C.E.

    1999-07-01

    Stirling-cycle engines have been identified as a promising technology for the conversion of concentrated solar energy into usable electrical power. In previous experimental work, the author has demonstrated that a heat pipe receiver can significantly improve system performance over a directly-illuminated heater head. The design and operating conditions of a heat pipe receiver differ significantly from typical laboratory heat pipes. New wick structures have been developed to exploit the characteristics of the solar generation system. Typically, these wick structures allow vapor generation within the wick. Conventional heat pipe models do not handle this enhancement, yet it can more than double the performance of the wick. In this study, the author developed a steady-state model of a boiling-enhanced wick for a solar heat pipe receiver. The model is used for design-point calculations and is written in FORTRAN90. Some limited comparisons have been made with actual test data.

  2. Nuclear piping system damping data studies

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. 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

  4. Structural integrity assessment of piping components

    International Nuclear Information System (INIS)

    Kushwaha, H.S.; Chattopadhyay, J.

    2008-01-01

    Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200-400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to component. The cyclic loading and system compliance effect on the load carrying capacity of piping components are investigated and new recommendations are made. (author)

  5. 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.

  6. 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

  7. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  8. Statistical investigation of the crack initiation lives of piping structural welded joint in low cycle fatigue test of 240 degree C

    International Nuclear Information System (INIS)

    Zhao Yongxiang; Gao Qing; Cai Lixun

    1999-01-01

    A statistical investigation into the fitting of four possible fatigue assumed distributions (three parameter Weibull, two parameter Weibull, lognormal and extreme maximum value distributions) for the crack initiation lives of piping structural welded joint in low cycle fatigue test of 240 degree C is performed by linear regression and least squares methods. The results reveal that the three parameters Weibull distribution may give misleading results in fatigue reliability analysis because the shape parameter is often less than 1. This means that the failure rate decreases with fatigue cycling which is contrary to the general understanding of the behaviour of welded joint. Reliability analyses may also affected by the slightly nonconservative evaluations in tail regions of this distribution. The other three distributions are slightly poor in the total fit effects, but they can be safety assumed in reliability analyses due to the non-conservative evaluations in tail regions mostly and the consistency with the fatigue physics of the structural behaviour of welded joint in the range of engineering practice. In addition, the extreme maximum value distribution is in good consists with the general physical understanding of the structural behaviour of welded joint

  9. 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

  10. Lab-scale impact test to investigate the pipe-soil interaction and comparative study to evaluate structural responses

    Directory of Open Access Journals (Sweden)

    Dong-Man Ryu

    2015-07-01

    Full Text Available This study examined the dynamic response of a subsea pipeline under an impact load to determine the effect of the seabed soil. A laboratory-scale soil-based pipeline impact test was carried out to investigate the pipeline deformation/strain as well as the interaction with the soil-pipeline. In addition, an impact test was simulated using the finite element technique, and the calculated strain was compared with the experimental results. During the simulation, the pipeline was described based on an elasto-plastic analysis, and the soil was modeled using the Mohr-Coulomb fail-ure criterion. The results obtained were compared with ASME D31.8, and the differences between the analysis results and the rules were specifically investigated. Modified ASME formulae were proposed to calculate the precise structural behavior of a subsea pipeline under an impact load when considering sand- and clay-based seabed soils.

  11. Cantilever beam test in Zr-2.5%Nb: Comparative study between pipe material obtained by laminating and wire drawing

    International Nuclear Information System (INIS)

    Gomez, A G; Szieber, W C; Vizcaino, P; Loureiro, N; Bianchi, D R; Banchik, A D

    2012-01-01

    The Zr-2.5Nb alloy is used in the manufacture of pressure tubes for the CANDU nuclear power reactors. These tubes are subjected to severe service conditions: one o f them, the heavy water corrosion due to the coolant generates release of hydrogen, part of which ingress in the material raising its initial concentration and exposing them to a phenomena referred as delay hydrogen cracking. The results presented in this paper show the performance of a pressure tube of domestic manufacture under conditions of tension and hydrogen content in order to be compared with the behavior of a standard pressure tube in operation in the nuclear power plant. To do this is, cantilever notched and pre cracked samples were hydrided from both kinds of tubes. Each one of these samples was subjected to the cantilever beam test, which consists in a bending test performed in a furnace at 250 o C. Starting from a stress intensity factor Ki which determines the propagation start of the crack, the growth is followed by the acoustic emission technique up to the arrest of the crack by controlling the bending load. This work presents the comparative data such as critical voltages, behavior of hydrides, and DHC parameters from both trials. Although the number of tests is reduced; results show a good performance of the tubes of domestic manufacture (author)

  12. Project no. 40 in the field of district heating. Testing method for casing thickness of preinsulated district heating pipes. Fjernvarmeprojekt nr. 40. Proevemetode for praeisolerede fjernvarmeroers kappetykkelser

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, K. E.; Hoej, N. P.

    1989-01-15

    The investigation evaluates the resistance of casing to penetration of a subject (eg. stones). Tests have been performed on cones with apex angles of 15 and 45 degrees, and on a mandrel with a diameter of 10mm. Results are expressed by the maximum force measured during penetration. Penetration velocities were 500 mm/min., 100mm./min. and 0.1 mmm/min. Casings with a thickness of 2.0mm-10.6mm, and diameters of 110mm and 125mm respectively, have been tested, and penetration tests were made on a plane plate. It is concluded that the greatest stresses are caused by a concentrated load and that resistance is proportionate to casing thickness. (Some material qualities for PUR cellular plastic and PEH casings are described). It has not been possible to find a connection between a casing's diameter and its resistance, but it is concluded that the strenght may be proportional to the decimal logarithm for the load velocity. Resistance to penetration of the mandrel (diam.10mm) is 278 X t (t=thickness of casing), of the cone 45 deg. is 145.7 x t and of the cone 15 deg. is 65.9 x t. - where the load velocity is 0.1mm/min. The strength of the casing is evaluated in proportion to the net weight of a preinsulated district heating pipe and to the influence under foaming. It appears that with a maximum very sharp-8mm-150 cone stone there is a risk of damage to the thinnest casing (2.2mm-3mm). If no stones appear with a cone sharper than 45 deg., there will be no risk from sizes below 8mm. However, safety evaluations should take special account of the casing's temperature. (AB).

  13. Pressure piping systems examination. 2. ed

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-01

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

  14. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  15. Radiation-resistance of polyurethane pipes for cooling liquid in BES III

    International Nuclear Information System (INIS)

    Li Xunfeng; Zheng Lifang; Ji Quan; Wu Ping; Wang Li

    2009-01-01

    Gamma ray radiation and neutron radiation are predominant in the working conditions of BES III, and the radiation-resistance aging of polyurethane pipes is very important in this condition, as the pipes of cooling liquid for beam pipe and SCQ (superconducting quadrupole) vacuum pipe in BESIII. Polyester polyurethane pipes and polyether polyurethane pipes were irradiated by gamma ray and neutron. The radiation doses were as much as ten years' doses in BES. Pressure test, FTIR and thermal analysis were used to study the radiation-resistance of these two kinds of polyurethane pipes. The results show that the radiation-resistance and thermal stability of polyester polyurethane pipes are prior to those of polyether polyurethane pipes, and the pressure resistance of polyester polyurethane pipes is almost maintained after the irradiation by gamma ray and neutron, but the polyether polyurethane pipes can be aged and ruptured after the irradiation by neutron. (authors)

  16. 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.

  17. Erosion and break-up of light-gas layers by a horizontal jet in a multi-vessel, large-scale containment test system

    Energy Technology Data Exchange (ETDEWEB)

    Zboray, Robert, E-mail: robert.zboray@psi.ch; Mignot, Guillaume; Kapulla, Ralf; Paladino, Domenico

    2015-09-15

    The distribution and eventual stratification of hydrogen released during a hypothetical severe accident and the stability of the stratification formed in the early phase of the transient is of particular safety concern in Light Water Reactors (LWRs). The large-scale containment test facility PANDA (PSI, Switzerland) has been used to perform a series of four tests examining the erosion and break-up of stratified light-gas layers in the frame of the OECD SETH-2 project. The ultimate goal of the test program is to set-up an experimental data base of high-quality and high-density data that can challenge and validate 3D containment codes like e.g. GOTHIC, GASFLOW or MARS and validate the applicability of CFD codes like FLUENT or CFX for LWR containment problems. The test series discussed here focuses on the erosion of a stratified, helium-rich layer by horizontal steam injection at different locations below the layer. An approach with step-wise increasing complexity has been chosen to examine this problem allowing control over the rate of pressure increase and the occurrence of condensation. The step-wise approach enables a thorough understanding of the influence of different phenomena like position of steam injection, diffusion, pressurization and condensation on the behavior and erosion of the stratified layer.

  18. Study of the performances of acoustic emission testing for glass fibre reinforced plastic pipes containing defects; Etude des performances du controle par emission acoustique de tubes en materiaux composites resine-fibre de verre contenant des defauts

    Energy Technology Data Exchange (ETDEWEB)

    Villard, D.; Vidal, M.C.

    1995-08-01

    Glass fibre reinforced plastic pipes are more and more often used, in nuclear power plants, for building or replacement of water pipings classified `nuclear safety`. Tests have been performed to evaluate the performances of acoustic emission testing for in service inspection of these components. The tests were focused on glass fibre reinforced polyester and vinyl-ester pipes, in as received conditions or containing impacts, and intentionally introduced defects. They have been carried out by CETIM, following the ASTM Standard E 1118 (code CARP), to a maximum pressure lever of 25 Bar The results show that the CARP procedure can be used for detection of defects and evaluation of their noxiousness towards internal pressure: most of the tubes containing low energy impacts could not be distinguished from tubes without defect; on the other hand the important noxiousness of lacks of impregnation of roving layer appeared clearly. Complementary tests have been performed on some tubes at a more important pressure lever, for which the damage of the tubes in enough to deteriorate there elastic properties. The results showed that CARP procedure give valuable informations on damage level. It would be interesting to evaluate acoustic emission on tubes containing realistic in-service degradations. (author). 11 refs., 15 figs., 6 tabs., 2 appends.

  19. 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.)

  20. 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.