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

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

1984-08-01

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

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

International Nuclear Information System (INIS)

1984-08-01

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

Statistical Dependence of Pipe Breaks on Explanatory Variables

Directory of Open Access Journals (Sweden)

Patricia Gómez-Martínez

2017-02-01

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

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)

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

Kaliatka, A.; Valinčius, M.

2012-01-01

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

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

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.)

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.

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

Directory of Open Access Journals (Sweden)

Massoud Tabesh

2011-07-01

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

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

International Nuclear Information System (INIS)

1989-01-01

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

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.

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)

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

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

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.

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)

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

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)

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.

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)

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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.)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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.)

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

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.

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)

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

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.

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)

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

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)

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

Goerner, F.; Munz, D.

1984-01-01

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

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.

Non-destructive qualification tests for ITER cryogenic axial insulating breaks

International Nuclear Information System (INIS)

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

2014-01-01

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

Non-destructive qualification tests for ITER cryogenic axial insulating breaks

Energy Technology Data Exchange (ETDEWEB)

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

2014-01-29

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

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.)

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.

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.

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

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.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

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.

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

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

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.))

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.

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)

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.)

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)

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)

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.)

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

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

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

Internal testing of pipe systems with IRIS inspection system

International Nuclear Information System (INIS)

1986-01-01

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

Fatigue 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

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

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)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

Nilsson, L.; Sjoeberg, A.

1987-01-01

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

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

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

International Nuclear Information System (INIS)

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

1986-12-01

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

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.)

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.)

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

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.

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)

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

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

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…

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.

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...

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-02-15

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

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

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

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)

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.

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

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)

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

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)

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.)

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

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

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

Uncertainty analysis for probabilistic pipe fracture evaluations in LBB applications

International Nuclear Information System (INIS)

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

1997-01-01

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

Uncertainty analysis for probabilistic pipe fracture evaluations in LBB applications

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

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...

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)

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)

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

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)

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

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

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)

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)

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

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

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

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.

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.

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

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

International Nuclear Information System (INIS)

Dunn, F.E.

1990-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

Theuer, E.; Heller, M.

1979-01-01

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

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.

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.

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

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.

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)

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

International Nuclear Information System (INIS)

Faidy, C.; Gilles, P.

1997-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

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)

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

International Nuclear Information System (INIS)

Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

2005-01-01

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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)

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)

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.

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

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.

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

A multi-step approach for evaluation of pipe impact effects

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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

Directory of Open Access Journals (Sweden)

Won-Hee Kang

2017-01-01

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

Conceptual design of pipe whip restraints using interactive computer analysis

International Nuclear Information System (INIS)

Rigamonti, G.; Dainora, J.

1975-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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)

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)

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

Probabilistic pipe fracture evaluations for leak-rate-detection applications

International Nuclear Information System (INIS)

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

1995-04-01

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

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

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.

Analysis of the SBLOCAs in HANARO pool for 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, Y. J.

2004-09-01

Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

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

International Nuclear Information System (INIS)

1985-04-01

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

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

African Journals Online (AJOL)

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

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

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

Indian Academy of Sciences (India)

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

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)

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)

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

Development of integrated insulation joint for cooling pipe in tokamak reactor

International Nuclear Information System (INIS)

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

1994-08-01

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

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

Analysis of th SBLOCAs in the room 1 for 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, Y. J.

2004-10-01

Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss of Coolant Accidents (SBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperature (PCT) is predicted to be about 931.4 .deg. C for the cold leg break accident in PWR fuel test mode and 931.6 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 8% of the cross section area of the pipe for PWR fuel test mode and the 10% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

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.)

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)

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)

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.)

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

Failure probability estimate of type 304 stainless steel piping

International Nuclear Information System (INIS)

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

1989-01-01

The primary source of in-service degradation of the SRS production reactor process water piping is intergranular stress corrosion cracking (IGSCC). IGSCC has occurred in a limited number of weld heat affected zones, areas known to be susceptible to IGSCC. A model has been developed to combine crack growth rates, crack size distributions, in-service examination reliability estimates and other considerations to estimate the pipe large-break frequency. This frequency estimates the probability that an IGSCC crack will initiate, escape detection by ultrasonic (UT) examination, and grow to instability prior to extending through-wall and being detected by the sensitive leak detection system. These events are combined as the product of four factors: (1) the probability that a given weld heat affected zone contains IGSCC; (2) the conditional probability, given the presence of IGSCC, that the cracking will escape detection during UT examination; (3) the conditional probability, given a crack escapes detection by UT, that it will not grow through-wall and be detected by leakage; (4) the conditional probability, given a crack is not detected by leakage, that it grows to instability prior to the next UT exam. These four factors estimate the occurrence of several conditions that must coexist in order for a crack to lead to a large break of the process water piping. When evaluated for the SRS production reactors, they produce an extremely low break frequency. The objective of this paper is to present the assumptions, methodology, results and conclusions of a probabilistic evaluation for the direct failure of the primary coolant piping resulting from normal operation and seismic loads. This evaluation was performed to support the ongoing PRA effort and to complement deterministic analyses addressing the credibility of a double-ended guillotine break

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-04-01

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

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

International Nuclear Information System (INIS)

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

1995-04-01

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

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

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

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

Application of LBB to high energy piping systems in operating PWR

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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.

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.

Vibration test on KMRR reactor structure and primary cooling system piping

Energy Technology Data Exchange (ETDEWEB)

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

1994-10-01

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

Vibration test on KMRR reactor structure and primary cooling system piping

International Nuclear Information System (INIS)

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

1994-10-01

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

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

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.

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

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

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

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

Construction of the two-phase critical flow test facility

International Nuclear Information System (INIS)

Chung, C. H.; Chang, S. K.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

2002-03-01

The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

Analysis of the test results for the two-phase critical flow with non-condensible gas

International Nuclear Information System (INIS)

Chang, S. K.; Chung, C. H.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

2002-07-01

The two-phase critical flow test was performed for simulating the pipe break accident of SMART reactor. The requirements of the critical flow test are 7∼20mm pipe break dia., 7∼12MPa stagnation pressure, 0∼60 .deg. C subcooling degree and 0∼0.5kg/s N 2 gas flow rate. The test section is sharp edged pipe type which has the dimension of I.D.=20, L=300mm and I.D.=10.9, L=1000mm. The test conditions are 4, 7, 10 MPa at stagnation pressure, 0, 20, 50 .deg. C of subcooling degree and 0.028∼0.39 kg/s of N 2 injection gas flowrate. The measured data at test section and other components in terms of pressure, temperature and flowrate were collected in DAS computer with maintaining the steady state conditions at least 60 seconds. From the test results, the critical characteristics of the break pipe were analysed and verified the capacity of the test facility. For the verification of the Modified Henry-Fauske model which can predict the two-phase critical flow with non-condensible gas, the code simulation using MARS which contains the option of the Modified Henry -Fauske model was performed. The simulation results of steady-state two-phase critical flow experiments show that they agree with the measured critical flow rates within 6% root-mean-square error

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.

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

International Nuclear Information System (INIS)

Julie M Jarvis; Allen T Vieira; James M Gilmer

2005-01-01

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

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.

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

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Design for whipping pipe impact on reinforced concrete panels

International Nuclear Information System (INIS)

Chen, C.C.; Gurbuz, O.

1984-01-01

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

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...

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)

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

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

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)

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)

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.

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)

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.

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

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.

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

International Nuclear Information System (INIS)

Bieselt, R.; Wolf, M.

1995-01-01

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

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

International Nuclear Information System (INIS)

Zahoor, A.; Gamble, R.

1984-01-01

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

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

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.

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.

Evaluation of J-integral estimation scheme for flawed throughwall pipes

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1987-02-01

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

Evaluation of J-integral estimation scheme for flawed throughwall pipes

International Nuclear Information System (INIS)

Zahoor, A.

1987-01-01

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

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.

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

International Nuclear Information System (INIS)

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

1997-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-06-01

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

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

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.

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.

Evaluation of flawed-pipe experiments: Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.

1986-11-01

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

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

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

Fatigue analysis of HANARO primary cooling system piping

International Nuclear Information System (INIS)

Ryu, Jeong Soo

1998-05-01

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

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.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

Ferng, Y.M.

2008-01-01

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

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.

Generic safety evaluation report regarding integrity of BWR scram system piping

International Nuclear Information System (INIS)

1981-08-01

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

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

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

International Nuclear Information System (INIS)

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

1994-07-01

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

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

International Nuclear Information System (INIS)

Jong, Rudolf Peter de

2004-01-01

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

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

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.

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.

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.

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

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

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%.

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

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

Energy Technology Data Exchange (ETDEWEB)

Kottmann, A

1978-01-01

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

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.

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)

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)

Development of two-phase flow along a large vertical pipe

International Nuclear Information System (INIS)

Dirk Lucas; Prasser, H.M.

2005-01-01

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

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

International Nuclear Information System (INIS)

Novotny, J.; Novotny, J. Jr.

2003-01-01

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

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

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

WHIPJET progress on piping restraint elimination at Beaver Valley - 2

International Nuclear Information System (INIS)

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

1986-01-01

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

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

Application of the cracked pipe element to creep crack growth prediction

Energy Technology Data Exchange (ETDEWEB)

Brochard, J.; Charras, T.

1997-04-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Reactor Materials Program probability of indirectly--induced failure of L and P reactor process water piping

International Nuclear Information System (INIS)

Daugherty, W.L.

1988-01-01

The design basis accident for the Savannah River Production Reactors is the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping material. The Reactor Materials Program was initiated to provide the technical basis for an alternate credible design basis accident. One aspect of this work is to determine the probability of the DEGB; to show that in addition to being incredible, it is also highly improbable. The probability of a DEGB is broken into two parts: failure by direct means, and indirectly-induced failure. Failure of the piping by direct means can only be postulated to occur if an undetected crack grows to the point of instability, causing a large pipe break. While this accident is not as severe as a DEGB, it provides a conservative upper bound on the probability of a direct DEGB of the piping. The second part of this evaluation calculates the probability of piping failure by indirect causes. Indirect failure of the piping can be triggered by an earthquake which causes other reactor components or the reactor building to fall on the piping or pull it from its supports. Since indirectly-induced failure of the piping will not always produce consequences as severe as a DEGB, this gives a conservative estimate of the probability of an indirectly- induced DEGB. This second part, indirectly-induced pipe failure, is the subject of this report. Failure by seismic loads in the piping itself will be covered in a separate report on failure by direct causes. This report provides a detailed evaluation of L reactor. A walkdown of P reactor and an analysis of the P reactor building provide the basis for extending the L reactor results to P reactor

Development of residual stress prediction model in pipe weldment

Energy Technology Data Exchange (ETDEWEB)

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

2002-03-15

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

Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow

International Nuclear Information System (INIS)

Kuzay, T.M.; Kasza, K.E.

1985-01-01

A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 90 0 sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions

Suitability of pipeline material for buried gas and water piping

Energy Technology Data Exchange (ETDEWEB)

Funk, R

1976-01-01

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

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

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

International Nuclear Information System (INIS)

Hofmann, E.; Heeschen, U.

1979-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Kishida, K.; Zahoor, A.

1988-08-01

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

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

International Nuclear Information System (INIS)

Kishida, K.; Zahoor, A.

1988-08-01

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

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.

Rupture hardware minimization in pressurized water reactor piping

International Nuclear Information System (INIS)

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

1989-01-01

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

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)

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-15

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

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

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

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

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...

Pipe inspection using the pipe crawler. Innovative technology summary report

International Nuclear Information System (INIS)

1999-05-01

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

Pipe inspection using the pipe crawler. Innovative technology summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-05-01

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

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Kuzay, T.M.; Kasza, K.E.

1985-01-01

A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 90/sup 0/ sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions.

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.

Experimental benchmark for piping system dynamic-response analyses

International Nuclear Information System (INIS)

1981-01-01

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

Experimental benchmark for piping system dynamic response analyses

International Nuclear Information System (INIS)

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

1981-01-01

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

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)

Butt-welding technology for double walled Polyethylene pipe

International Nuclear Information System (INIS)

Lee, Bo-Young; Kim, Jae-Seong; Lee, Sang-Yul; Kim, Yeong K.

2012-01-01

Highlights: ► We developed a butt welding apparatus for doubled walled Polyethylene pipe. ► We design the welding process by analyzing thermal behaviors of the material. ► We performed the welding and tested the welded structural performances. ► We also applied the same technology to PVC pipes. ► We verified the butt welding was successful and effective for the pipes with irregular sections. -- Abstract: In this study, mechanical analyses of a butt welding technology for joining Polyethylene pipe are presented. The pipe had unique structure with double wall, and its section topology was not flat. For an effective repair of leakage and replacements of the pipe, the butt welding technology was developed and tested. For the material characterizations, thermodynamic analyses such as thermal gravimetric analysis and differential scanning calorimetry were performed. Based on the test results, the process temperature and time were determined to ensure safe joining of the pipes using a hot plate apparatus. The welding process was carefully monitored by measuring the temperature. Then, the joined pipes were tested by various methods to evaluate the quality. The analyses results showed the detail process mechanism during the joining process, and the test results demonstrated the successful application of the technology to the sewage pipe repairs.

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

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

International Nuclear Information System (INIS)

Streit, R.D.

1981-01-01

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

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)

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.

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

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...

Tensile and fracture properties of primary heat transport system piping material

International Nuclear Information System (INIS)

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

1997-07-01

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

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.

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

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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.)

Integrity assessment of the cold leg piping system in a PWR

International Nuclear Information System (INIS)

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

1981-01-01

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

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.

Pipe rupture hardware minimization in pressurized water reactor system

International Nuclear Information System (INIS)

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

1987-01-01

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

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.

Performance correlations for high temperature potassium heat pipes

International Nuclear Information System (INIS)

Merrigan, M.A.; Keddy, E.S.; Sena, J.T.

1987-01-01

Potassium heat pipes designed for operation at a nominal temperature of 775K have been developed for use in a heat pipe cooled reactor design. The heat pipes operate in a gravity assist mode with a maximum required power throughput of approximately 16 kW per heat pipe. Based on a series of sub-scale experiments with 2.12 and 3.2 cm diameter heat pipes the prototypic heat pipe diameter was set at 5.7 cm with a simple knurled wall wick used in the interests of mechanical simplicity. The performance levels required for this design had been demonstrated in prior work with gutter assisted wicks and emphasis in the present work was on the attainment of similar performance with a simplified wick structure. The wick structure used in the experiment consisted of a pattern of knurled grooves in the internal wall of the heat pipe. The knurl depth required for the planned heat pipe performance was determined by scaling of wick characteristic data from the sub-scale tests. These tests indicated that the maximum performance limits of the test heat pipes did not follow normal entrainment limit predictions for textured wall gravity assist heat pipes. Test data was therefore scaled to the prototype design based on the assumption that the performance was controlled by an entrainment parameter based on the liquid flow depth in the groove structure. This correlation provided a reasonable fit to the sub-scale test data and was used in scale up of the design from the 8.0 cm 2 cross section of the largest sub-scale heat pipe to the 25.5 cm 2 cross section prototype. Correlation of the model predictions with test data from the prototype is discussed

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

Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test)

International Nuclear Information System (INIS)

1977-06-01

Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations

Early results of gate valve flow interruption blowdown tests

International Nuclear Information System (INIS)

DeWall, K.G.

1988-01-01

The preliminary results of the USNRC/INEL high-energy BWR line break flow interruption testing are presented. Two representative nuclear valve assemblies were cycled under design basis Reactor Water Cleanup pipe break conditions to provide input for the technical basis for resolving the Nuclear Regulatory Commission's Generic Issue 87. The effects of the blowdown hydraulic loadings on valve operability, especially valve closure stem forces, were studied. The blowdown tests showed that, given enough thrust, typical gate valves will close against the high flow resulting from a line break. The tests also showed that proper operator sizing depends on the correct identification of values for the sizing equation. Evidence exists that values used in the past may not be conservative for all valve applications. The tests showed that improper operator lock ring installation following test or maintenance can invalidate in-situ test results and prevent the valve from performing its design function. 2 refs., 12 figs., 2 tabs

Evaluation of piping fracture analysis method by benchmark study, 1

International Nuclear Information System (INIS)

Takahashi, Yukio; Kashima, Koichi; Kuwabara, Kazuo

1987-01-01

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

INEL/USNRC pipe damping experiments and studies

International Nuclear Information System (INIS)

Ware, A.G.

1987-08-01

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

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

International Nuclear Information System (INIS)

Lowry, W.

1994-01-01

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Norris, D.M.

1984-11-01

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

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

International Nuclear Information System (INIS)

Zahoor, A.; Norris, D.M.

1984-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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.

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.

Ductile fracture of circumferentially cracked pipes subjected to bending loads

International Nuclear Information System (INIS)

Zahoor, A.; Kanninen, M.F.

1981-01-01

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

Ductile fracture of circumferentially cracked pipes subjected to bending loads

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Kanninen, M.F.

1981-10-01

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

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

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 ...

Base-plate effects on pipe-support stiffness

International Nuclear Information System (INIS)

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

1981-01-01

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

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

International Nuclear Information System (INIS)

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

1997-01-01

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

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

International Nuclear Information System (INIS)

Verezemskij, V.G.

1986-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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.

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.)

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

Experimental basis for parameters contributing to energy dissipation in piping systems

International Nuclear Information System (INIS)

Ibanez, P.; Ware, A.G.

1985-01-01

The paper reviews several pipe testing programs to suggest the phenomena causing energy dissipation in piping systems. Such phenomena include material damping, plasticity, collision in gaps and between pipes, water dynamics, insulation straining, coupling slippage, restraints (snubbers, struts, etc.), and pipe/structure interaction. These observations are supported by a large experimental data base. Data are available from in-situ and laboratory tests (pipe diameters up to about 20 inches, response levels from milli-g's to responses causing yielding, and from excitation wave forms including sinusoid, snapback, random, and seismic). A variety of pipe configurations have been tested, including simple, bare, straight sections and complex lines with bends, snubbers, struts, and insulation. Tests have been performed with and without water and at zero to operating pressure. Both light water reactor and LMFBR piping have been tested

Further considerations for damping in heavily insulated pipe systems

International Nuclear Information System (INIS)

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

1985-01-01

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

Further considerations for damping in heavily insulated pipe systems

International Nuclear Information System (INIS)

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

1985-01-01

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

Design of a cavity heat pipe receiver experiment

Science.gov (United States)

Schneider, Michael G.; Brege, Mark H.; Greenlee, William J.

1992-01-01

A cavity heat pipe experiment has been designed to test the critical issues involved with incorporating thermal energy storage canisters into a heat pipe. The experiment is a replication of the operation of a heat receiver for a Brayton solar dynamic power cycle. The heat receiver is composed of a cylindrical receptor surface and an annular heat pipe with thermal energy storage canisters and gaseous working fluid heat exchanger tubes surrounding it. Hardware for the cavity heat pipe experiment will consist of a sector of the heat pipe, complete with gas tube and thermal energy storage canisters. Thermal cycling tests will be performed on the heat pipe sector to simulate the normal energy charge/discharge cycle of the receiver in a spacecraft application.

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.)

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

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

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

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)

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

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

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.

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.

Load-deflection characteristics of small bore insulated pipe clamps

International Nuclear Information System (INIS)

Severud, L.K.; Clark, G.L.

1982-01-01

High temperature LMFBR piping is subject to rapid temperature changes during transient events. Typically, this pipe is supported by specially designed insulated pipe clamps to prevent excessive thermal stress from developing during these transients. The special insulated clamps used on both FFTF and CRBR piping utilize a Belleville spring arrangement to compensate for pipe thermal expansion. Analysis indicates that this produces a non-linear, directionally sensitive clamp spring rate. Since these spring rates influence the seismic response of a supported piping system, it was deemed necessary to evaluate them further by test. This has been accomplished for the FFTF clamps. A more standard insulated pipe clamp, which does not incorporate Belleville springs to accommodate thermal expansion, was also tested. This type clamp is simple in design, and economically attractive. It may have wide application prospects for use in LMFBR small bore auxiliary piping operating at temperatures below 427 0 C. Load deflection tests were conducted on 2.54 CM and 7.62 CM diameter samples of these commercial clamps

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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.

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

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.

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.

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

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)

Fabrication of lithium/C-103 alloy heat pipes for sharp leading edge cooling

Science.gov (United States)

Ai, Bangcheng; Chen, Siyuan; Yu, Jijun; Lu, Qin; Han, Hantao; Hu, Longfei

2018-05-01

In this study, lithium/C-103 alloys heat pipes are proposed for sharp leading edge cooling. Three models of lithium/C-103 alloy heat pipes were fabricated. And their startup properties were tested by radiant heat tests and aerothermal tests. It is found that the startup temperature of lithium heat pipe was about 860 °C. At 1000 °C radiant heat tests, the operating temperature of lithium/C-103 alloy heat pipe is lower than 860 °C. Thus, startup failure occurs. At 1100 °C radiant heat tests and aerothermal tests, the operating temperature of lithium/C-103 alloy heat pipe is higher than 860 °C, and the heat pipe starts up successfully. The startup of lithium/C-103 alloy heat pipe decreases the leading edge temperature effectively, which endows itself good ablation resistance. After radiant heat tests and aerothermal tests, all the heat pipe models are severely oxidized because of the C-103 poor oxidation resistance. Therefore, protective coatings are required for further applications of lithium/C-103 alloy heat pipes.

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

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

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

International Nuclear Information System (INIS)

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

1997-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-01

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

Effect on mechanical properties of glass reinforced epoxy (GRE) pipe filled with different geopolymer filler molarity for piping application

Science.gov (United States)

Hashim, M. F. Abu; Abdullah, M. M. A.; Ghazali, C. M. R.; Hussin, K.; Binhussain, M.

2017-04-01

This study investigated the use of a novel white clay geopolymer as a filler to produce high strength glass reinforced epoxy pipe. It was found that using white clay geopolymer as filler gives better compressive strength to the glass reinforced epoxy pipe. The disadvantages of current glass reinforced epoxy pipes such low compressive strength which can be replaced by the composite pipes. Geopolymerization is an innovative technology that can transform several aluminosilicate materials into useful products called geopolymers or inorganic polymers. A series of glass reinforced epoxy pipe and glass reinforced epoxy pipe filled with 10 - 40 weight percentages white clay geopolymer filler with 4 Molarity and 8 Molarity were prepared. Morphology of white clay geopolymer filler surface was indicates using scanning electron microscopy. The additions of white clay geopolymer filler for both 4 Molarity and 8 Molarity show higher compressive strength than glass reinforced epoxy pipe without any geopolymer filler. The compressive test of these epoxy geopolymer pipe samples was determined using Instron Universal Testing under compression mode. Nonetheless, the compressive strength of glass reinforced epoxy pipe with white clay geopolymer filler continues to drop when added to 40 wt% of the geopolymer filler loading for both 4 Molarity and 8 Molarity. These outcomes showed that the mixing of geopolymer materials in epoxy system can be attained in this research.

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

International Nuclear Information System (INIS)

Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

2000-01-01

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

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Criteria for accepting piping vibrations measured during FFTF plant startup