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Sample records for primary pipe rupture

  1. 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.)

  2. Analysis of the FFTF primary pipe rupture transients

    International Nuclear Information System (INIS)

    Perkins, K.R.; Bari, R.A.; Chen, L.C.; Albright, D.C.

    1979-01-01

    The response of the Fast Flux Test Facility (FFTF) to hypothetical ruptures of the high pressure primary piping has been analyzed using two LMFBR plant systems codes, namely IANUS and DEMO. Comparisons of the average channel temperatures predicted by the two codes show good agreement for identical transients. However, the hot channel temperatures predicted by DEMO are about 60K higher than the corresponding IANUS predictions for severe transients. This difference is attributed to the dynamic hot channel factors employed in DEMO which discount the thermal inertia of the duct walls for rapid transients. DEMO also predicts more severe transients for hot-leg ruptures in FFTF than previously reported analyses for the CRBR

  3. Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albright, D.C.; Bari, R.A.

    1976-04-01

    In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident.

  4. Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Albright, D.C.; Bari, R.A.

    1976-04-01

    In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident

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

    International Nuclear Information System (INIS)

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    1975-10-01

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

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

    International Nuclear Information System (INIS)

    2004-08-01

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

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

    International Nuclear Information System (INIS)

    1975-03-01

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

  9. Primary coolant pipe rupture study AT(49-24)-0202

    International Nuclear Information System (INIS)

    Hale, D.A.; Clarke, W.L. Jr.

    1977-01-01

    Fatigue crack growth rate tests were conducted on 304 stainless steel and 516 carbon steel in a simulated BWR primary water environment. A study was carried out to determine the feasibility of measuring sensitization in type 304 SS by use of an Electrochemical Potentiokinetic Reactivation (EPR) technique, develop correlations between degree of sensitization (as measured electrochemically) and the intergranular stress corrosion cracking (IGSCC) resistance of type 304 SS, and provide technical data for evaluating the degree of sensitization and IGSCC susceptibility of welded components. 27 figures, 8 tables

  10. Air ingress behavior during a primary-pipe rupture accident of HTGR

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki

    1997-11-01

    The inherent properties of a HTGR facilitates the design with high degree of passive safe performances, compared to other type. However, it is still not clear if the present HTGR can maintain a passive safe function during a primary-pipe rupture accident, or what would be design criteria to guarantee the HTGR with the high degree of passive safe performances during the accident. To investigate safe characteristics, the study has been performed experimentally and analytically on the air ingress behavior during the accident. It was indicated that there are two stages in the accident of the HTGR having a reverse U-shaped channel. In the first stage, an air ingress process limits molecular diffusion and natural circulation of the gas mixture having a very slow velocity. In the second stage, the air ingress process limits the ordinary natural circulation of air throughout the reactor. A numerical calculation code has been developed to analyze thermal-hydraulic behavior during the first stage. This code provides a numerical method for analyzing a transport phenomena in a multi-component gas system by solving one-dimensional basic equations and using a flow network model. It was possible to predict or analyze the air ingress process regarding the density of the gas mixture, concentration of each gas species and duration of the first stage of the accident. It was indicated that the safe characteristics of the HTGR from the present experiment as follows. The safety cooling rate that the air ingress process terminates during the first stage exists in the HTGR having the reverse U-shaped channel. Moreover, the ordinary natural circulation of air can not produce in the second stage by injecting helium from the bottom of the pressure vessel corresponding the low-temperature side channel. Therefore, it was found that the idea of helium injection is one of useful methods for the prevention of air ingress and of graphite corrosion in the future HTGRs. (J.P.N.). 74 refs

  11. Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

    1991-11-01

    When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

  12. Impact analyses after pipe rupture

    International Nuclear Information System (INIS)

    Chun, R.C.; Chuang, T.Y.

    1983-01-01

    Two of the French pipe whip experiments are reproduced with the computer code WIPS. The WIPS results are in good agreement with the experimental data and the French computer code TEDEL. This justifies the use of its pipe element in conjunction with its U-bar element in a simplified method of impact analyses

  13. Secondary pipe rupture at Mihama unit 3

    International Nuclear Information System (INIS)

    Hajime Ito; Takehiko Sera

    2005-01-01

    The secondary system pipe rupture occurred on August 9, 2004, while Mihama unit 3 was operating at the rated thermal power. The rupture took place on the condensate line-A piping between the No.4 LP heater and the deaerator, downstream of an orifice used for measuring the condensate flux. The pipe is made of carbon steel, and normally has 558.8 mm diameter and 10 mm thickness. The pipe wall had thinned to 0.4 mm at the point of minimum thickness. It is estimated that the disturbed flow of water downstream of the orifice caused erosion/corrosion and developed wall thinning, leading to a rupture at the thinnest section under internal pressure, about 1MPa. Observation of the pipe internal surface revealed a scale-like pattern typical in this kind of phenomenon. Eleven workers who were preparing for an annual outage that was to start from August 14 suffered burn injuries, of who five died. Since around 1975, we, Kansai Electric, have been checking pipe wall thickness while focusing on the thinning of carbon steel piping in the secondary system. Summarizing the results from such investigation and reviewing the latest technical knowledge including operating experience from overseas utilities, we compiled the pipe thickness management guideline for PWR secondary pipes, 1990. The pipe section that ruptured at the Mihama unit 3 should have been included within the inspection scopes according to the guideline but was not registered on the inspection list. It had not been corrected for almost thirty years. As the result, this pipe section had not been inspected even once since the beginning of the plant operation, 1976. It seems that the quality assurance and maintenance management had not functioned well regarding the secondary system piping management, although we were responsible for the safety of nuclear power plants as licensee. We will review the secondary system inspection procedure and also improve the pipe thickness management guideline. And also, we would replace

  14. Rupture disc opening property for using pipe rupture test in JAERI

    International Nuclear Information System (INIS)

    Kato, Rokuro

    1983-03-01

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

  15. Study on air ingress during an early stage of a primary-pipe rupture accident of a high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Hishida, M.; Takeda, T.

    1991-01-01

    A primary-pipe rupture accident is one of the design-based accidents of the HTTR. As the first step of our final goal of predicting the multicomponent gas flow in a reactor during the early stages of the accident, the present paper aims at studying experimentally and analytically, the basic features of air ingress and gas transportation by transient molecular diffusion and the transient natural convection of a two-component gas mixture. The present paper comprises two main parts. The first part deals with analytical and experimental studies on N 2 ingress (corresponding to air ingress) and gas transportation by molecular diffusion and the one-dimensional natural convection of an He-N 2 two-component gas mixture in a reverse-U-shaped tube. Analytical and experimental results are discussed on the N 2 mole fraction change with time after the simulated pipe rupture and on the initation time of the natural circulation of pure N 2 . The second part deals with a preliminary simulation test of air ingress during the early stages of the accident. The test is performed with a very simple model of the reactor. The experimental results are discussed on the change in mole fraction of air with time and on the initiation time of the natural circulation of pure air. (orig.)

  16. Failure analysis on a ruptured petrochemical pipe

    Energy Technology Data Exchange (ETDEWEB)

    Harun, Mohd [Industrial Technology Division, Malaysian Nuclear Agency, Ministry of Science, Technology and Innovation Malaysia, Bangi, Kajang, Selangor (Malaysia); Shamsudin, Shaiful Rizam; Kamardin, A. [Univ. Malaysia Perlis, Jejawi, Arau (Malaysia). School of Materials Engineering

    2010-08-15

    The failure took place on a welded elbow pipe which exhibited a catastrophic transverse rupture. The failure was located on the welding HAZ region, parallel to the welding path. Branching cracks were detected at the edge of the rupture area. Deposits of corrosion products were also spotted. The optical microscope analysis showed the presence of transgranular failures which were related to the stress corrosion cracking (SCC) and were predominantly caused by the welding residual stress. The significant difference in hardness between the welded area and the pipe confirmed the findings. Moreover, the failure was also caused by the low Mo content in the stainless steel pipe which was detected by means of spark emission spectrometer. (orig.)

  17. Safety design guide for pipe rupture protection for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for pipe rupture protection identifies high-energy systems in which pipe ruptures must be postulated to occur, as well as systems that must be protected from the dynamic effects of such ruptures. Dynamic effects considered in this SDG consist of pipe whip (including missiles generated by pipe ruptures, if any) and jet impingement, Requirements for protection against the dynamic effects of a postulated pipe rupture and method of protection of essential structures, systems and components are specified for these effects. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 5 refs. (Author) .new

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

    International Nuclear Information System (INIS)

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

    1983-02-01

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

  19. 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

  20. 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

  1. Instationary discharge rates and shear factors in pipe ruptures

    International Nuclear Information System (INIS)

    Pana, P.

    1976-01-01

    The loads observed in ruptures of steam- or water-conducting pipes may occur as reactive forces on the pipes themselves or as jet forces on the structural components adjacent to the point of rupture. The present paper deals with the instationary acceleration phase directly after rupture. The general laws of conservation (mass, energy, momentum) may be used, but in their instationary form. This results in a system of partial differential equations which does not provide a comprehensive mathematical solution. However, since efficient electronic computer systems are available, difference methods are increasingly often used. Such calculations were carried out for water-steam as an ideal gas and under simplifying assumptions. (orig./AK) [de

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  3. Estimative of core damage frequency in IPEN's IEA-R1 research reactor (PSA level 1) due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami

    2009-01-01

    This work applies the methodology of probabilistic safety assessment level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid by major pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, emergency core cooling system (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  4. Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

    2009-01-01

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

  5. Full instantaneous traversal rupture of the primary loop pipeline

    International Nuclear Information System (INIS)

    Baytelesov, S.A.; Kungurov, F.R.

    2010-01-01

    Accident, reflecting full immediate cross rupture of primary loop pipe of WWR-SM research reactor of INP AS RUz is observed in this paper. Calculations for accident situation and analysis for different reactor cores, formed from fully IRT-3M type high enriched fuel (36% enrichment on 235 U), first mixed core, compiled from 16 IRT-3M fuel assemblies and 4 IRT-4M type fuel assemblies with low enriched fuel (19,7% enrichment on 235 U) and the core fully formed from low enriched fuel are carried out

  6. Acoustic system for pipe rupture monitoring and leak detection

    International Nuclear Information System (INIS)

    Herzog, W.; Jonas, H.

    1982-06-01

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

  7. Report of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    2001-12-01

    In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The following findings and conclusion were made as the result of the present examination. (1) Wall thickness of the pipe was significantly reduced in the ruptured region. (2) Dimple pattern resulting from ductile fracture by shearing was observed in the fracture surfaces of nearly all of the pieces and no indication of fatigue crack growth was found. (3) Microstructure showed a typical carbon

  8. Correlation of energy balance method to dynamic pipe rupture analysis

    International Nuclear Information System (INIS)

    Kuo, H.H.; Durkee, M.

    1983-01-01

    When using an energy balance approach in the design of pipe rupture restraints for nuclear power plants, the NRC specifies in its Standard Review Plan 3.6.2 that the input energy to the system must be multiplied by a factor of 1.1 unless a lower value can be justified. Since the energy balance method is already quite conservative, an across-the-board use of 1.1 to amplify the energy input appears unneccessary. The paper's purpose is to show that this 'correlation factor' could be substantially less than unity if certain design parameters are met. In this paper, result of nonlinear dynamic analyses were compared to the results of the corresponding analyses based on the energy balance method which assumes constant blowdown forces and rigid plastic material properties. The appropriate correlation factors required to match the energy balance results with the dynamic analyses results were correlated to design parameters such as restraint location from the break, yield strength of the energy absorbing component, and the restraint gap. It is shown that the correlation factor is related to a single nondimensional design parameter and can be limited to a value below unity if appropriate design parameters are chosen. It is also shown that the deformation of the restraints can be related to dimensionless system parameters. This, therefore, allows the maximum restraint deformation to be evaluated directly for design purposes. (orig.)

  9. 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)

  10. Data book of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    2002-03-01

    In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The result of the present examination has already been reported to NISA and has also been published as the JAERI-Tech report No.2001-94. This report is a data book containing the detailed data obtained by the present examination. (author)

  11. 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

  12. Simulation of the behaviour of a servo actuated check valve upon rupture of the feedwater pipe

    International Nuclear Information System (INIS)

    Lucas, A.M. de; Perezagua, R.L.; Rosa, B. de la; Sanz, J.

    1995-01-01

    The steam generator replacement programme at Almaraz NPP, provides for the installation of a replacement damped non-return valve for the feedwater system. the function of this valve is to protect the steam generator in the event of a rupture in the feedwater pipe. Sudden closure of the check valve, against the flow and following rupture of the feedwater pipe, causes overpressure in the valve which is transmitted to the steam generator nozzle. It is therefore necessary to know this when designing the internal systems of the steam generator. Using the RELAP5/MODE3 code, it has been possible to simulate the dynamic behaviour of a check valve upon rupture of a feedwater pipe postulated outside the containment. The calculation model has been applied to different types of check valve. (Author)

  13. Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

    International Nuclear Information System (INIS)

    Lu, S.C.; Larder, R.A.; Ma, S.M.

    1981-01-01

    This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

  14. 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.

  15. 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

  16. Application of discrete scale invariance method on pipe rupture

    International Nuclear Information System (INIS)

    Rajkovic, M.; Mihailovic, Z.; Riznic, J.

    2007-01-01

    'Full text:' A process of material failure of a mechanical system in the form of cracks and microcracks, a catastrophic phenomenon of considerable technological and scientific importance, may be forecasted according to the recent advances in the theory of critical phenomena in statistical physics. Critical rupture scenario states that, in many concrete and composite heterogeneous materials under compression and materials with large distributed residual stresses, rupture is a genuine critical point, i.e., the culmination of a self-organization of damage and cracking characterized by power law signatures. The concept of discrete scale invariance leads to a complex critical exponent (or dimension) and may occur spontaneously in systems and materials developing rupture. It establishes, theoretically, the power law dependence of a measurable observable, such as the rate of acoustic emissions radiated during loading or rate of heat released during the process, upon the time to failure. However, the problem is the power law can be distinguished from other parametric functional forms such as an exponential only close to the critical time. In this paper we modify the functional renormalization method to include the noise elimination procedure and dimension reduction. The aim is to obtain the prediction of the critical rupture time only from the knowledge of the power law parameters at early times prior to rupture, and based on the assumption that the dynamics close to rupture is governed by the power law dependence of the temperature measured along the perimeter of the tube upon the time-to-failure. Such an analysis would not only enhance the precision of prediction related to the rupture mechanism but also significantly help in determining and predicting the leak rates. The prediction will be compared to experimental data on Zr-2.5%Nb made tubes. Note: The views expressed in the paper are those of the authors and do not necessary represents those of the commission. (author)

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

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki

    1985-01-01

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

  18. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

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

    International Nuclear Information System (INIS)

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

    1982-09-01

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

  20. Delayed primary realignment of posterior urethral rupture | Shittu ...

    African Journals Online (AJOL)

    The treatment of acute posterior urethral rupture is controversial. Twelve patients who presented with acute posterior urethral rupture over a five--year period were treated by delayed primary realignment of the injury. The technique of this procedure and the outcome are the subject of this presentation. Eight patients had ...

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

    International Nuclear Information System (INIS)

    Weiss, G.

    1987-01-01

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

  2. Pressurization of a compartment due to the rupture of coolant piping

    International Nuclear Information System (INIS)

    Kot, C.A.; Hsieh, B.J.

    1993-01-01

    The pressurization and venting of enclosed compartments due to the accidental rupture of coolant piping is a safety problem common to many nuclear facilities. The processes associated with such an accident are very complex, involving, in general, transient multiphase flows, interactions and mixing between the incoming flows and the gases in the compartment, and heat transfer with the surroundings. Since pipe rupture is associated with many phenomenological uncertainties, elaborate 3-D thermal-hydraulic modeling and extensive calculational efforts are not warranted for many design applications. It is then more appropriate to rely. on simplified, global analysis approaches which can provide reasonably conservative estimates of the structural loads and flow processes, and which can readily be used in parameter/design studies. The objective of this paper is to present such an approach

  3. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

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

  4. Thermal aging of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.; Brenner, S.S.; Spitznagel, J.A.

    1985-01-01

    The long term mechanical integrity of the pipes used to carry the primary cooling water in a pressurized water nuclear reactor is of the utmost importance for safe operation. A combined atom probe field-ion microscopy (APFIM) and transmission electron microscopy (TEM) study was performed to characterize the microstructure of this cast stainless steel and to determine the changes that occur during long-term low-temperature thermal aging. The material used in this investigation was a commercial CF 8 type stainless. The steel was examined in the as-cast, unaged condition and also after aging for 7500 h at 673K. 3 refs., 4 figs., 2 tabs

  5. Design considerations for the protection from the effects of pipe rupture

    International Nuclear Information System (INIS)

    1975-11-01

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

  6. 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)

  7. 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.

  8. Study on heat transfer and fluid flow in the stand pipe rupture accident

    International Nuclear Information System (INIS)

    Fumizawa, Motoo; Hishida, Makoto

    1991-09-01

    This paper deals with an experimental investigation of the buoyancy driven exchange flow which takes place through a narrow cylindrical channel, during the stand pipe rupture accident in a high temperature gas-cooled reactor (HTGR). The velocity distribution through the cylindrical channel is measured by a laser Doppler velocimeter, in order to evaluate the air ingress flow rate. The experiments are performed under atmospheric pressure with nitrogen as a working fluid. Rayleigh number ranges from 1.3 x 10 7 to 7.0 x 10 7 . The following conclusions were obtained: (1) The laser Doppler velocimeter was found a good method for the measurement of the velocity of the exchange flow. (2) When the temperature of the hemisphere and the bottom heated plate, which simulate the top cover of the reactor, was kept uniform, the volumetric exchange flow rate agreed well with Epstein's result. (3) The exchange flow through a narrow cylindrical channel fluctuated irregularly with time and space. (author)

  9. An extended risk assessment approach for chemical plants applied to a study related to pipe ruptures

    International Nuclear Information System (INIS)

    Milazzo, Maria Francesca; Aven, Terje

    2012-01-01

    Risk assessments and Quantitative Risk Assessment (QRA) in particular have been used in the chemical industry for many years to support decision-making on the choice of arrangements and measures associated with chemical processes, transportation and storage of dangerous substances. The assessments have been founded on a risk perspective seeing risk as a function of frequency of events (probability) and associated consequences. In this paper we point to the need for extending this approach to place a stronger emphasis on uncertainties. A recently developed risk framework designed to better reflect such uncertainties is presented and applied to a chemical plant and specifically the analysis of accidental events related to the rupture of pipes. Two different ways of implementing the framework are presented, one based on the introduction of probability models and one without. The differences between the standard approach and the extended approaches are discussed from a theoretical point of view as well as from a practical risk analyst perspective.

  10. Refurbishment of the IEAR1 primary coolant system piping supports

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  11. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years

  12. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  13. 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

  14. Weathering effects on tensile and stress rupture strength of glass fiber reinforced vinylester and epoxy thermoset pipes

    Science.gov (United States)

    Nizamuddin, Syed

    Glass fiber reinforced vinylester (GFRE) and epoxy (GFRE) pipes have been used for more than three decades to mitigate corrosion problems in oil fields, chemical and industrial plants. In these services, both GFRV and GFRE pipes are exposed to various environmental conditions. Long-term mechanical durability of these pipes after exposure to environmental conditions, which include natural weathering exposure to seasonal temperature variation, sea water, humidity and other corrosive fluids like crude oil, should be well known. Although extensive research has been undertaken, several major issues pertaining to the performance of these pipes under a number of environmental conditions still remain unresolved. The main objective of this study is to investigate the effects of natural weathering, combined natural weathering with seawater and crude oil exposure, for time periods ranging from 3 to 36 months respectively, on the tensile and stress rupture behavior of GFRV and GFRE pipes. Ring specimens are machined from GFRV and GFRE pipes and tested before and after exposure to different weathering conditions prevalent in the eastern region (Dhahran) of Saudi Arabia and present under service conditions. The natural weathering and combined natural weathering with crude oil exposure of GFRV specimens revealed increased tensile strength even after 36 months of exposure when compared with that of the as received samples. However, the combined natural weathering with seawater exposure of GFRV samples revealed better tensile behavior till 24 months of exposure, and after 36 months their tensile strength was seen to be below that of the as received GFRV samples. The stress rupture behavior of natural weather exposed GFRV samples showed an improvement after 12 months of exposure and it decreased after 24 and 36 months of exposure when compared with the as received GFRV samples. The combined natural weathering with crude oil and seawater exposure of GFRV sample revealed improved

  15. 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

  16. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  17. Response of the primary piping loop to an HCDA

    International Nuclear Information System (INIS)

    Chang, Y.W.; Moneim, M.T.A.; Wang, C.Y.; Gvildys, J.

    1975-01-01

    The paper describes a method for analyzing the response of the primary piping loop that consists of straight pipes, elbows, and other components connected in series and subject to hypothetical core disruptive accident (HCDA) loads at both ends of the loop. The complete hydrodynamic equations in two-dimensions, that include both the nonlinear convective and viscous dissipation terms are used for the fluid dynamics together with the implicit ICE technique. The external walls of the pipes and components are treated as thin shells in which the analysis accounts for the membrane and bending strength of the wall, elastic-plastic material behavior, as well as large deformation under the effect of transient loading conditions. In the straight pipes, the flow is assumed to be axisymmetric; in the elbow regions, the two dimensions considered are the r and theta directions. The flow in the other components is also assumed to be axisymmetric; the components are modeled as a circular cylinder, in which the radius of the cylinder can be varied to conform with the outside shape of the component and the flow area inside can be changed independently from the outside shape. However, they must remain axially symmetric. The method is applied to a piping loop which consists of six elastic-plastic pipes and five rigid elbows connected in series and subjected to pressure pulses at both ends of the loop

  18. Numerical Analysis on the Compressible Flow Characteristics of Supersonic Jet Caused by High-Pressure Pipe Rupture Using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong-Kil; Yoon, Jun-Kyu [Gachon Univ., Sungnam (Korea, Republic of); Kim, Kwang-Chu [KEPCO-E& C, Kimchun (Korea, Republic of)

    2017-10-15

    A rupture in a high-pressure pipe causes the fluid in the pipe to be discharged in the atmosphere at a high speed resulting in a supersonic jet that generates the compressible flow. This supersonic jet may display complicated and unsteady behavior in general . In this study, Computational Fluid Dynamics (CFD) analysis was performed to investigate the compressible flow generated by a supersonic jet ejected from a high-pressure pipe. A Shear Stress Transport (SST) turbulence model was selected to analyze the unsteady nature of the flow, which depends upon the various gases as well as the diameter of the pipe. In the CFD analysis, the basic boundary conditions were assumed to be as follows: pipe of diameter 10 cm, jet pressure ratio of 5, and an inlet gas temperature of 300 K. During the analysis, the behavior of the shockwave generated by a supersonic jet was observed and it was found that the blast wave was generated indirectly. The pressure wave characteristics of hydrogen gas, which possesses the smallest molecular mass, showed the shortest distance to the safety zone. There were no significant difference observed for nitrogen gas, air, and oxygen gas, which have similar molecular mass. In addition, an increase in the diameter of the pipe resulted in the ejected impact caused by the increased flow rate to become larger and the zone of jet influence to extend further.

  19. 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

  20. Assessment of value-impact associated with the elimination of postulated pipe ruptures from the design basis for nuclear power plants

    International Nuclear Information System (INIS)

    Holman, G.S.; Chou, C.K.

    1985-01-01

    The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping systems is also assessed in a less quantitative manner

  1. Successful primary repair of late diagnosed spontaneous esophageal rupture: A case report

    Directory of Open Access Journals (Sweden)

    Diana Y. Kircheva

    2017-01-01

    Conclusion: Primary closure of late diagnosed spontaneous esophageal rupture can be successful, even when it is complicated by a prolonged delay in treatment and failed endoscopic procedures. We conclude that primary surgical repair should be attempted in patients with spontaneous esophageal rupture if tissues are viable.

  2. IEA-R1 renewed primary coolant piping system stress analysis

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. 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)

  4. Esophageal Rupture as a Primary Manifestation in Eosinophilic Esophagitis

    Directory of Open Access Journals (Sweden)

    Natalia Vernon

    2014-01-01

    Full Text Available Eosinophilic esophagitis (EoE is a chronic inflammatory process characterized by symptoms of esophageal dysfunction and, histologically, by eosinophilic infiltration of the esophagus. In adults, it commonly presents with dysphagia, food impaction, and chest or abdominal pain. Chronic inflammation can lead to diffuse narrowing of the esophageal lumen which may cause food impaction. Endoscopic procedures to relieve food impaction may lead to complications such as esophageal perforation due to the friability of the esophageal mucosa. Spontaneous transmural esophageal rupture, also known as Boerhaave’s syndrome, as a primary manifestation of EoE is rare. In this paper, we present two adult patients who presented with esophageal perforation as the initial manifestation of EoE. This rare complication of EoE has been documented in 13 other reports (11 adults, 2 children and only 1 of the patients had been previously diagnosed with EoE. A history of dysphagia was present in 1 of our patients and in the majority of previously documented patients. Esophageal perforation is a potentially severe complication of EoE. Patients with a history of dysphagia and patients with spontaneous esophageal perforation should warrant an evaluation for EoE.

  5. Rupture

    CERN Multimedia

    Association du personnel

    2006-01-01

    Our Director-General is indifferent to the tradition of concertation foreseen in our statutes and is "culturally" unable to associate the Staff Association with problem-solving in staff matters. He drags his heels as long as possible before entering into negotiations, presents "often misleading" solutions at the last minute which he only accepts to change once a power struggle has been established. Faced with this rupture and despite its commitment to concertation between gentlemen. The results of the poll in which the staff is invited to participate this week. We therefore need your support to state our claims to the Governing Bodies. The Staff Association proposes a new medium of communication and thus hopes to show that it is ready for future negotiations. The pages devoted to the Staff Association are presented in a more informative, reactive and factual manner and in line with the evolution of the social situation at CERN. We want to establish strong and continuous ties between the members of CERN and ou...

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

    International Nuclear Information System (INIS)

    Jarman, B.L.

    1985-10-01

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

  7. Development of forging technology for PWR primary piping

    International Nuclear Information System (INIS)

    Morin, F.; Badeau, J.P.; Lambs, R.

    1996-01-01

    The purpose of this presentation is to give information on the changes in the design and manufacture of Primary Piping for electronuclear boilers of the Pressurized Water Reactor type (PWR) which has resulted in the making of one-piece forged lines including stub pipes and arcs. The optimization of these items is aimed at improving the life of the new power stations as well as guaranteeing their safety, while reducing inspection and maintenance requirements in service. The demonstration of the manufacturing feasibility has just been completed. It has taken material form in the installation, on the CIVAUX 1 section, of the first one-piece cold leg in the world. It will shortly be followed by the installation on the CIVAUX 2 section of a complete loop of bent forged pipes. Therefore, this new know-how is going to be incorporated in the French Rules (RCC-M) and can be directly taken into consideration both in the next work to be done and in the design and definition of a future nuclear reactor

  8. 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.

  9. Removal of Shippingport Station primary system components and piping

    International Nuclear Information System (INIS)

    LaGuardia, T.S.; Lipsett, S.M.

    1987-01-01

    The dismantling workscope for the Shippingport Station Decommissioning Project was divided into subtasks to permit the work to be subcontracted to the maximum extent practicable. Major subtasks were identified and described by Activity specifications which could then be grouped into logical work packages to be put out for bid. Two of the largest dismantling work packages, removal of piping and components, were grouped together and designated as Activity Specifications 4 and 5. TLG Services, Inc. and Cleveland Wrecking Company formed a Joint Venture to perform this work during a two-year period at a cost of approximately $7 million. The major portions of this dismantling workscope are described. The primary system components within this workscope consist of the stainless steel reactor coolant piping, check valves, reactor coolant pumps, steam generators, and reactor purification demineralizers and coolers. The work performed, the heavy rigging preparations and procedures, the cutting tools used, component draining/capping techniques to prevent spills, contamination containment, airborne control techniques, and lessons learned during the removal of these primary system components are described. Summaries of crew size and composition, labor hours, duration hours and radiation exposure to workers are provided and discussed briefly. The successful completion of this work is evidence of the engineering, planning, equipment, materials and labor pool available to remove large, radioactively contaminated components safely. This experience will help decommissioning planners to prepare for the removal of reactor components in future decommissioning

  10. Long-term creep rupture strength of weldment of Fe-Ni based alloy as candidate tube and pipe for advanced USC boilers

    Energy Technology Data Exchange (ETDEWEB)

    Bao, Gang; Sato, Takashi [Babcok-Hitachi K.K., Hiroshima (Japan). Kure Research Laboratory; Marumoto, Yoshihide [Babcok-Hitachi K.K., Hiroshima (Japan). Kure Div.

    2010-07-01

    A lot of works have been going to develop 700C USC power plant in Europe and Japan. High strength Ni based alloys such as Alloy 617, Alloy 740 and Alloy 263 were the candidates for boiler tube and pipe in Europe, and Fe-Ni based alloy HR6W (45Ni-24Fe-23Cr-7W-Ti) is also a candidate for tube and pipe in Japan. One of the Key issues to achieve 700 C boilers is the welding process of these alloys. Authors investigated the weldability and the long-term creep rupture strength of HR6W tube. The weldments were investigated metallurgically to find proper welding procedure and creep rupture tests are ongoing exceed 38,000 hours. The long-term creep rupture strengths of the HST weld joints are similar to those of parent metals and integrity of the weldments was confirmed based on with other mechanical testing results. (orig.)

  11. Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

    Science.gov (United States)

    Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

    2012-10-01

    Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

  12. Configuration analysis of pipe support for primary cooling using Ps + Caepipe code

    International Nuclear Information System (INIS)

    Sitandung, Y. B.; Pustandyo, W.; Sujalmo, S.

    1998-01-01

    Pipe stress evaluation and support loads has been analyzed on piping segment of RSG-GAS primary cooling system. This paper describes an analysis method of piping system with the use of computer Code PS + CAEPIPE Version 3.4.05.W. From the selected pipe segment, the data of pipe characteristic, material properties, operation condition, equipment and supports were used input. The final evaluation result of primary cooling pipe segment show that actual stress dead weight and seismic load are less than allowable limits (stress ratio 0.101 for deadweight 0.35 for seismic load). From the above ratio, it can be concluded that ratio of pipe support configuration to stress distribution is acceptable, and based on analysis result, the Code used by INTERATOM was sufficiently accurate

  13. Temperature distribution in the Temelin NPP primary circuit piping

    International Nuclear Information System (INIS)

    Blaha, V.; Maca, K.; Kodl, P.; Kroj, L.

    2004-01-01

    Temperature non-homogeneity in the VVER 1000 reactor primary piping hot legs was detected during the commissioning of Temelin units 1 and 2. A quantification of temperature differences was carried out and explanation of its causes was presented. Mathematical analysis of the effect was carried out using the PHOENICS 3.4 code, and the results were processed graphically by means of a post processor PHOTON and by means of a user program allowing statistic evaluation of temperature profiles at the core outlet and in the area of the temperature-measurement pits. The coolant temperatures in the core area increased gradually following the given radial and axial distribution of output from the inlet temperature of 288.1 degC to 315-331 degC at the core outlet. The temperature profile was balanced and in the IO piping in the area of temperature-measurement pits the difference of the maximum and minimum temperature value was approx. 1 degC according to the calculation. The temperature field shape is mainly determined by the radial distribution of the core output. The mean outlet temperature from the core weighted through mass flow is determined by the flow through the core and by the total output. The calculated temperature span at the core outlet in the range of 315 - 331 degC corresponded well with the measured values during the operation. The values were in the range of 310-333 degC, however, the in-core thermocouple inaccuracy should also be taken into consideration. On the other hand, the temperature span in the area of temperature-measurement pits was actually about 4 times higher than the calculated temperature (observed: 4 degC as against the calculated 1 degC). A good agreement was reached between the analysis results and the actual condition of the nuclear unit in the area of the core outlet. (P.A.)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  15. A rare cause of pleural effusion: ruptured primary pleural hydatid cyst.

    Science.gov (United States)

    Erkoç, Mustafa Fatih; Öztoprak, Bilge; Alkan, Sevil; Okur, Aylin

    2014-03-06

    Hydatidosis is an endemic parasitic disease in Mediterranean countries, often caused by the dog tapeworm Echinococcus granulosus. The disease predominantly affects the liver (60-70%) and lungs (30%), and the surgical management is considered as the gold standard for treatment. Besides anaphylactic reactions, the most frequent complication of the hydatid disease is rupture into neighbouring structures, often affecting the bronchi, gastrointestinal tract and peritoneal/pleural cavities, according to its location. Primary pleural hydatidosis is an extremely rare entity and we present a ruptured pleural hydatid cyst with unusual location.

  16. Investigation of the conservatism associated with different combinations between primary and secondary piping responses

    International Nuclear Information System (INIS)

    Wang, Y.K.; Subudhi, M.; Bezler, P.

    1983-01-01

    This report includes the findings of an investigation of the conservatism associated with different combinations between the primary and secondary stress components for piping systems under dynamic loading, such as in an earthquake event. The primary stresses are induced by piping response to its mass inertia effects. The secondary stresses are induced by relative displacements of piping supports. The study involves an independnent time history analysis of several typical piping models to predict a best estimate of the actual dynamic and pseudo-static pipe responses to an earthquake. These piping systems are also analyzed using the response spectrum method to obtain the maximum primary stress components. Secondary stresses are next calculated by performing a set of static analyses which provide the worst stress condition. The two components are then combined by both SRSS and absolute sum methods as the results are compared with time history solutions. It is found that the SRSS combination of the primary and secondary stress components yield acceptable results provided the secondary stress component is calculated in the most unfavorable phasing relationship among displacements of piping supports

  17. The integral analysis of 40 mm diameter pipe rupture in cooling system of fusion facility W7-X with ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Kačegavičius, Tomas, E-mail: Tomas.Kacegavicius@lei.lt; Povilaitis, Mantas, E-mail: Mantas.Povilaitis@lei.lt

    2015-12-15

    Highlights: • The analysis of loss-of-coolant accident (LOCA) in W7-X facility. • Burst disc is sufficient to prevent pressure inside the plasma vessel exceeding 110 kPa. • Developed model of the cooling system adequately represents the expected phenomena. - Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental facility of stellarator type, which is currently being built at the Max-Planck-Institute for Plasmaphysics located in Greifswald, Germany. W7-X shall demonstrate that in future the energy could be produced in such type of fusion reactors. The safety analysis is required before the operation of the facility could be started. A rupture of 40 mm diameter pipe, which is connected to the divertor unit (module for plasma cooling) to ensure heat removal from the vacuum vessel in case of no-plasma operation mode “baking” is one of the design basis accidents to be investigated. During “baking” mode the vacuum vessel structures and working fluid – water are heated to the temperature 160 °C. This accident was selected for the detailed analysis using integral code ASTEC, which is developed by IRSN (France) and GRS mbH (Germany). This paper presents the integral analysis of W7-X response to a selected accident scenario. The model of the main cooling circuit and “baking” circuit was developed for ASTEC code. There were analysed two cases: (1) rupture of a pipe connected to the upper divertor unit and (2) rupture of a pipe connected to the lower divertor unit. The results of analysis showed that in both cases the water is almost completely released from the units into the plasma vessel. In both cases the pressure in the plasma vessel rapidly increases and in 28 s the set point for burst disc opening is reached preventing further pressurisation.

  18. Mechanical Properties of Post Irradiation Primary Cooling Piping of Bandung Research Reactor

    International Nuclear Information System (INIS)

    Histori; Renaningsih S; Sri Nitiswati; Ari Triyadi

    2003-01-01

    Testing on primary coolant piping of research reactor Bandung have been done. Primary coolant piping were made from Al 6061-T6. The goal of this activity is to investigate the mechanical properties changes caused by aging process after 33 years in irradiated. Type of testing i.e visual examination, thickness measurement, tensile and hardness test were done. The test data shown that there was a deposit at the inside surface of pipe, thickness decreased about 0.2 mm, tensile strength is 293 MPa, yield strength is 262 MPa, while the hardness is about 83 HRE (mean value). The test data than compared with ASTM standard. As the conclusion tensile and yield strength of pipe still fulfill the ASTM requirements, except the hardness is unsignificantly less/decreased. (author)

  19. Rupture of a high pressure gas or steam pipe in a tunnel: a preliminary investigation of the jet thrust exerted on a tunnel barrier

    International Nuclear Information System (INIS)

    Baum, M.R.

    1988-04-01

    On power plant, if a high pressure pipe containing high temperature gas or steam were to rupture, sensitive equipment necessary for safety shutdown of the plant could possibly be incapacitated if exposed to the subsequent high temperature environment. In many plant configurations the high pressure pipework is contained in tunnels where it is possible to construct barriers which isolate one section of the plant from another, thereby restricting the spread of the high temperature fluid/air mixture. This paper describes a preliminary experimental investigation of the magnitude of the thrust likely to be exerted on such barriers by a gas jet issuing from the failed pipe. Measurements of the thrust exerted on a flat plate by normal impingement of a highly underexpanded gas jet are in agreement with a semi-quantitative analysis assuming conservation of the axial momentum of the jet. (author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  1. Team collaborative innovation management based on primary pipes automatic welding project

    International Nuclear Information System (INIS)

    Li Jing; Wang Dong; Zhang Ke

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Rahardjo, H.P.

    2011-01-01

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

  3. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  4. Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

    International Nuclear Information System (INIS)

    Yoder, Graydon L. Jr.; Harvey, Karen; Ferrada, Juan J.

    2011-01-01

    A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

  5. Stress analysis of primary pipe rigid support of the in pile loop

    International Nuclear Information System (INIS)

    Hasibuan, Dj.

    1998-01-01

    Base on requirement of the safety analysis report and operation planning preparation on the in pile loop by using the fuel bundle in the test section, the stress analysis of primary pipe support has been done. The analysis was performed for the 3 (three) points of pipe support, which are chosen by random selection, i.e.: GU 2001, GU 2002, and GU 2331. The analysis result showed that the maximum allowable stress was greater then the actual stress. It is concluded that the existing supports fulfil the safety requirement

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  7. Grain size control method for the nozzles of AP1000 primary coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shenglong [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Sun, Yanhui [Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Yang, Bin, E-mail: byang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Zhang, Mingxian [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China)

    2017-04-01

    Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

  8. Grain size control method for the nozzles of AP1000 primary coolant pipes

    International Nuclear Information System (INIS)

    Wang, Shenglong; Sun, Yanhui; Yang, Bin; Zhang, Mingxian

    2017-01-01

    Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

  9. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  10. 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.

  11. 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

  12. Model for cobalt 60/58 deposition on primary coolant piping in a boiling water reactor

    International Nuclear Information System (INIS)

    Dehollander, W.R.

    1979-01-01

    A first principles model for deposition of radioactive metals into the corrosion films of primary coolant piping is proposed. It is shown that the predominant mechanism is the inclusion of the radioactive species such as Cobalt 60 into the spinel structure of the corrosion film during the act of active corrosion. This deposition can occupy only a defined fraction of the available plus 2 valence sites of the spinel. For cobalt ions, this ratio is roughly 4.6 x 10 -3 of the total iron sites. Since no distinction is made between Cobalt 60, Cobalt 58, and Cobalt 59 in this process, the radioactivity associated with this inclusion is a function of the ratio of the radioactive species to the nonradioactive species in the water causing the corrosion of the pipe metal. The other controlling parameter is the corrosion rate of the pipe material. This can be a function of time, for example, and it shown that freshly descaled metal when exposed to the cobalt containing water can incorporate as much as 10 x 10 -3 cobalt ions per iron atom in the initial corrosion period. This has implications for the problem of decontaminating nuclear reactor piping. Equations and selected observations are presented without reference to any specifically identified reactor or utility, so as to protect any proprietary interest

  13. Piping Stress analysis for primary system of nuclear power plant AP-600

    International Nuclear Information System (INIS)

    Tjahjono, Hendro; Arhatari, B.D.; W, Pustandyo; Sitandung, J.B; Sudarmaji, Djoko

    1999-01-01

    Piping stress analysis for AP-600 primary system has been done using software CAEPIPE and PS-CAEPIPE. The loading applied to the system are static and seismic category I and II piping in reactor building have been analysed, those are PXS-900, CVS-110, PCS-030, CAS-700 and CCS-050. These system contain pipes with the normal diameter of 1 , 2 , 4 a nd 8 . The design pressures are in the range of 150oF to 300oF. The acceleration taken as input in PS-CAEPIPE is based on seismic response spectra of floor the piping is located. In CAEPIPE, the acceleration taken from the peak of response spectra multiplied by 1.7 all of the acceleration in this case are no more than 0.36g. The result shows that after locating some supports, all system are acceptable without snubbers. The maximum stress are 11210 psi for deadweight load and 35593 psi for total load (the allowable values are 15000 psi and 45000 psi). The maximum displacement are 0.123 in for deadweight load, 1.474 in for hot load seismic load (the allowable values are 0.125 in for deadweight and 2.5 in for total load). The difference results of the both software is mainly in seismic calculation where mare parameters can be evaluated by PS-CAEPIPE including to evaluate valves acceleration in seismic condition

  14. 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).

  15. 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)

  16. 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

  17. A computational procedure for the investigation of whipping effect on ITER High Energy Piping and its application to the ITER divertor primary heat transfer system

    Energy Technology Data Exchange (ETDEWEB)

    Spagnuolo, G.A., E-mail: Alessandro.Spagnuolo@kraftanlagen.com [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany); Dell’Orco, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Mazzei, M. [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany)

    2015-10-15

    Highlights: • High Energy Piping (HEP) are components containing water or steam with P ≥ 2.0 MPa and/or T ≥ 100 °C. • The whipping effect in HEP may cause dangerous domino effect with relative rupture propagation. • The rapture is envisaged or postulated according to the stress state of piping. • A FEM analysis is performed in order to study the dynamic of whipping effect. • Study of special support to avoid and/or mitigate the whipping effect. - Abstract: The Tokamak Cooling Water System of nuclear facility has the function to remove heat from plasma facing components maintaining coolant temperatures, pressures and flow rates as required and, depending on thermal-hydraulic requirements, its systems are defined as High Energy Piping (HEP) because they contain fluids, such as water or steam, at a pressure greater than or equal to 2.0 MPa and/or at a temperature greater than or equal to 100 °C, or even gas at pressure above the atmospheric one. The French standards contemplate the need to consider the whipping effect on HEP design. This effect happens when, after a double ended guillotine break, the reaction force could create a displacement of the piping which might affect adjacent components. A research campaign has been performed, in cooperation by ITER Organization and University of Palermo, to outline the procedure to check whether whipping effect might occur and assess its potential damage effects so to allow their mitigation. This procedure is based on the guidelines issued by U.S. Nuclear Regulatory Commission. The proposed procedure has been applied to the analysis of the whipping effect of divertor primary heat transfer system HEP, using a theoretical–computational approach based on the finite element method.

  18. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  19. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  20. LOFT transient thermal analysis for 10 inch primary coolant blowdown piping weld

    International Nuclear Information System (INIS)

    Howell, S.K.

    1978-01-01

    A flaw in a weld in the 10 inch primary coolant blowdown piping was discovered by LOFT personnel. As a result of this, a thermal analysis and fracture mechanics analysis was requested by LOFT personnel. The weld and pipe section were analyzed for a complete thermal cycle, heatup and Loss of Coolant Experiment (LOCE), using COUPLE/MOD2, a two-dimensional finite element heat conduction code. The finite element representation used in this analysis was generated by the Applied Mechanics Branch. The record of nodal temperatures for the entire transient was written on tape VSN=T9N054, and has been forwarded to the Applied Mechanics Branch for use in their mechanical analysis. Specific details and assumptions used in this analysis are found in appropriate sections of this report

  1. Numerical simulation of flows in a circular pipe transversely subjected to a localized impulsive body force with applications to blunt traumatic aortic rupture

    Energy Technology Data Exchange (ETDEWEB)

    Labbio, G Di; Keshavarz-Motamed, Z; Kadem, L, E-mail: lcfd@encs.concordia.ca [Department of Mechanical and Industrial Engineering, Concordia University, Montreal, Quebec, H3G 1M8 (Canada)

    2017-06-15

    Much debate surrounds the mechanisms responsible for the occurrence of blunt traumatic aortic rupture in car accidents, particularly on the role of the inertial body force experienced by the blood due to the abrupt deceleration. The isolated influence of such body forces acting on even simple fluid flows is a fundamental problem in fluid dynamics that has not been thoroughly investigated. This study numerically investigates the fundamental physical problem, where the pulsatile flow in a straight circular pipe is subjected to a transverse body force on a localized volume of fluid. The body force is applied as a brief rectangular impulse in three distinct cases, namely during the accelerating, peak, and decelerating phases of the pulsatile flow. A dimensionless number, termed the degree of influence of the body force (Ψ), is devised to quantify the relative strength of the body force over the flow inertia. The impact induces counter-rotating cross-stream vortices at the boundaries of the forced section accompanied by complex secondary flow structures. This secondary flow is found to develop slowest for an impact occurring during an accelerating flow and fastest during a decelerating flow. The peak skewness of the velocity field, however, occurred at successively later times for the three respective cases. After the impact, these secondary flows act to restore the unforced state and such dominant spatial structures are revealed by proper orthogonal decomposition of the velocity field. This work presents a new class of problems that requires further theoretical and experimental investigation. (paper)

  2. Numerical simulation of flows in a circular pipe transversely subjected to a localized impulsive body force with applications to blunt traumatic aortic rupture

    Science.gov (United States)

    Di Labbio, G.; Keshavarz-Motamed, Z.; Kadem, L.

    2017-06-01

    Much debate surrounds the mechanisms responsible for the occurrence of blunt traumatic aortic rupture in car accidents, particularly on the role of the inertial body force experienced by the blood due to the abrupt deceleration. The isolated influence of such body forces acting on even simple fluid flows is a fundamental problem in fluid dynamics that has not been thoroughly investigated. This study numerically investigates the fundamental physical problem, where the pulsatile flow in a straight circular pipe is subjected to a transverse body force on a localized volume of fluid. The body force is applied as a brief rectangular impulse in three distinct cases, namely during the accelerating, peak, and decelerating phases of the pulsatile flow. A dimensionless number, termed the degree of influence of the body force (Ψ), is devised to quantify the relative strength of the body force over the flow inertia. The impact induces counter-rotating cross-stream vortices at the boundaries of the forced section accompanied by complex secondary flow structures. This secondary flow is found to develop slowest for an impact occurring during an accelerating flow and fastest during a decelerating flow. The peak skewness of the velocity field, however, occurred at successively later times for the three respective cases. After the impact, these secondary flows act to restore the unforced state and such dominant spatial structures are revealed by proper orthogonal decomposition of the velocity field. This work presents a new class of problems that requires further theoretical and experimental investigation.

  3. 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

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

    International Nuclear Information System (INIS)

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

    1976-04-01

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

  5. 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)

  6. 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)

  7. On the failure probability of the primary piping of the PWR

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    International Nuclear Information System (INIS)

    Hahn, G.; Hoeft, E.

    1980-01-01

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

  9. Primary and Aggregate Size Distributions of PM in Tail Pipe Emissions form Diesel Engines

    Science.gov (United States)

    Arai, Masataka; Amagai, Kenji; Nakaji, Takayuki; Hayashi, Shinji

    Particulate matter (PM) emission exhausted from diesel engine should be reduced to keep the clean air environment. PM emission was considered that it consisted of coarse and aggregate particles, and nuclei-mode particles of which diameter was less than 50nm. However the detail characteristics about these particles of the PM were still unknown and they were needed for more physically accurate measurement and more effective reduction of exhaust PM emission. In this study, the size distributions of solid particles in PM emission were reported. PMs in the tail-pipe emission were sampled from three type diesel engines. Sampled PM was chemically treated to separate the solid carbon fraction from other fractions such as soluble organic fraction (SOF). The electron microscopic and optical-manual size measurement procedures were used to determine the size distribution of primary particles those were formed through coagulation process from nuclei-mode particles and consisted in aggregate particles. The centrifugal sedimentation method was applied to measure the Stokes diameter of dry-soot. Aerodynamic diameters of nano and aggregate particles were measured with scanning mobility particle sizer (SMPS). The peak aggregate diameters detected by SMPS were fallen in the same size regime as the Stokes diameter of dry-soot. Both of primary and Stokes diameters of dry-soot decreased with increases of engine speed and excess air ratio. Also, the effects of fuel properties and engine types on primary and aggregate particle diameters were discussed.

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

    International Nuclear Information System (INIS)

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

    1981-08-01

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

  11. 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.

  12. 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.

  13. 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

  14. Measurement of tritium activity in the aluminum pipe of JRR-2 heavy water primary cooling system using imaging plate

    International Nuclear Information System (INIS)

    Motoishi, Shoji; Kobayashi, Katsutoshi

    2000-12-01

    JRR-2 is the heavy water cooling type nuclear reactor, which has been operated for 36 years (1960-1976) and in the process of decommissioning at present. For this reason, evaluation of tritium quantity permeated into the pipe and apparatus of the primary coolant heavy water circulating system is important. In the Radioisotope Production Division, activity of tritium in aluminum pipe was measured with imaging plate (IP), liquid scintillation analyzer and high purity germanium detector (HPGe). After acrylic paints was applied for the region except for tritium contamination on the surface of aluminum pipe, only the oxidized contaminated part was dissolved by 1.5%(1.21M) HF for 3 minutes, and measured with IP. As a result, the tritium was found to permeate in the depth of 25 μm. Moreover, 90% of it was found to be distributed within 7 μm. (author)

  15. Residual-stresses in austenitic stainless-steel primary coolant pipes and welds of pressurized-water reactors

    International Nuclear Information System (INIS)

    Faure, F.; Leggatt, R.H.

    1996-01-01

    Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between -250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching -300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa. (author)

  16. Defect occurrence, detection, location and characterization; essential variables of the LBB concept application to primary piping

    Energy Technology Data Exchange (ETDEWEB)

    Crutzen, S.; Koble, T.D.; Lemaitre, P. [and others

    1997-04-01

    Applications of the Leak Before Break (LBB) concept involve the knowledge of flaw presence and characteristics. In Service Inspection is given the responsibility of detecting flaws of a determined importance to locate them precisely and to classify them in broad families. Often LBB concepts application imply the knowledge of flaw characteristics such as through wall depth; length at the inner diameter (ID) or outer diameter (OD) surface; orientation or tilt and skew angles; branching; surface roughness; opening or width; crack tip aspect. Besides detection and characterization, LBB evaluations consider important the fact that a crack could be in the weld material or in the base material or in the heat affected zone. Cracks in tee junctions, in homogenous simple welds and in elbows are not considered in the same way. Essential variables of a flaw or defect are illustrated, and examples of flaws found in primary piping as reported by plant operators or service vendors are given. If such flaw variables are important in the applications of LBB concepts, essential is then the knowledge of the performance achievable by NDE techniques, during an ISI, in detecting such flaws, in locating them and in correctly evaluating their characteristics.

  17. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

    1995-12-31

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

  18. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    International Nuclear Information System (INIS)

    Szczurek, J.

    1995-01-01

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open

  19. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Szczurek, J [Inst. of Atomic Energy, Swierk (Poland)

    1996-12-31

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

  20. 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)

  1. Initial primary endovascular treatment in the management of ruptured intracranial aneurysms: a prospective consecutive series

    International Nuclear Information System (INIS)

    Mejdoubi, Mehdi; Cognard, Christophe; Gigaud, Michel; Tremoulet, Michel; Albucher, Jean-Francois

    2006-01-01

    From January 1998 to December 2002, endovascular treatment (EVT) was used as first intention in all patients with ruptured aneurysms. The objective of this study was to analyze the results of this therapeutic strategy. Among 401 patients admitted with a subarachnoid hemorrhage (SAH), 73 (18%) had a nonaneurysmal perimesencephalic SAH, 28 were not explored by angiography due to very poor clinical status, and 28 with aneurysmal SAH were not treated due to poor clinical status. Thus, of the 300 patients with a proven aneurysmal SAH, 272 (83%) were treated. EVT was attempted in 230 patients and was successful in 222 (82%), and clipping was performed in 50 (18%). Finally, EVT was successful in 234 aneurysms (96.7%) in 222 patients out of 242 aneurysms in 230 patients (some of the patients were treated for more than one aneurysm in the same procedure). EVT-related morbidity occurred in ten patients (4.5%) and mortality in eight (3.6%). Rate of dependency or death (modified Rankin scale 3-5) was 24.5% at 26 months. Initially, complete aneurysm occlusion was obtained in 81%, a dog ear in 3.4%, a neck remnant in 8% and incomplete occlusion in 8.1% of the patients. At follow-up (mean 26 months), the occlusion rate remained stable at 75%. This consecutive prospective series shows that EVT can be performed routinely as first-intention treatment in most aneurysmal SAH. Using this therapeutic strategy, EVT was performed in 82% of patients with long-term clinical results similar to those of the ISAT study. (orig.)

  2. 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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  4. Characterisation of girth pipe weld for primary heat transport system of pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Singh, P.K.; Vaze, K.K.; Kushwaha, H.S.

    2002-01-01

    The weld and heat affected zone (HAZ) associated with the girth weld are most vulnerable regions of the piping system. The different regions of the weld joint such as the weld metal, HAZ and base metal lead to heterogeneous mechanical and metallurgical properties of the joints. Due to their different metallurgical and mechanical properties, the amounts of damage produced in these regions are different when the component is subjected to service condition. Thus, it is imperative to know the characteristics of these regions of a pipe weld in order to identify the weakest zone for safe designing of high energy piping components. In view of this necessity the present study has been planned to carry out complete characterisation of the weld joint of SA 333 Gr.6 steel pipe, in terms of its metallurgical, mechanical and fracture properties. The mechanical and fracture mechanics properties of the base metal, weld deposit and HAZ have been compared and correlated with reference to their microstructures. Weld joints of SA 333 Gr.6 steel pipe have been prepared by using GTAW root pass and SMAW filling of V-grove as per recommended welding procedure specifications (WPS) conforming to ASME Sec IX commonly used to fabricate nuclear piping system components. The emphasis of the study is to characterise base, weld and HAZ of the pipe weld in terms of chemical, metallurgical, mechanical and fracture mechanics properties. The fracture toughness behaviour of the welds and HAZ has been characterised by J-integral parameters. The fatigue crack growth rate has been characterised by Paris Law. Stretched zone width (SZW) has been measured under SEM to evaluate initiation fracture toughness. The estimated initiation fracture toughness based on SZW and blunting line given by EGF recommendation have been compared. The fracture mechanics properties of base, weld and HAZ has been determined and compared. The fracture mechanics properties of the weld and HAZ have been correlated to their

  5. High cyclic fatigue of PWR primary piping generated by the pressure pulsations in coolant

    International Nuclear Information System (INIS)

    Zd'arek, J.; Pecinka, L.; Zeman, V.

    1999-01-01

    The protection of nuclear piping Class 1, 2 and 3 against fatigue failure is according to standard western practise and is based on - determining the cumulative usage factor (CUF) using equation (11) of ASME Code, Section III, Article NB 3653 for Class 1 piping; - Markl experiments and equation (10) of ASME Code, Section III, Article NC/ND 3653 for Class 2/3 piping. These evaluations cover only low cyclic loading and the possible influence of high cyclic loading as for example vibratory stresses generated by the main circulating pumps are not taken into account. This problem is fully covered in the Czech and Russian codes. The goal of this paper is 1. to clarify the basic principles; 2. to discuss in detail the methodology for the calculation of high frequency vibratory stresses; and 3. to demonstrate with a numerical example, the degree of influence of the CUF. (orig.)

  6. US NRC research on the integrity of piping in nuclear reactor primary systems

    International Nuclear Information System (INIS)

    Serpan, C.Z. Jr.

    1983-01-01

    This paper has attempted to provide a ''snapshot'' of the activities underway in NRC on the subject of LWR piping integrity as of the summer and fall of 1983. The paper is necessarily vague on certain topics of policy because they are either under review or are under development and the outcome cannot be accurately forecast at this time. Particularly in the area of BWR pipe cracking, events are very rapid so that positions and actions described in this paper may well be obsolete by the time it is published. Nevertheless, the activities and positions are as accurate as possible at the time of writing. Certainly the longer-range aspects of the research program represent the current direction and intent of NRC; nevertheless, as results come in and actions occur in the licensing and regulation arena of operating reactors, the emphasis of the research programs will necessarily shift to accommodate them so as to remain as relevant as possible. Thus, this paper is useful to show the intentions of NRC in the area of research for LWR piping, and it is also useful to document the status of the regulations on piping for which the research is being performed. (orig.)

  7. Shippingport Station Decommissioning Project: Removal of piping and equipment and removal of primary system components

    International Nuclear Information System (INIS)

    1989-01-01

    This report is a technical synopsis of the removal of contaminated and non-contaminated piping and equipment from the Shippingport Station Decommissioning Project (SSDP). The information is provided as a part of the Technology Transfer Program to document dismantling activities in support of reactor decommissioning. 5 refs., 29 figs., 4 tabs

  8. Volumetric and chemical control auxiliary circuit for a PWR primary circuit

    International Nuclear Information System (INIS)

    Costes, D.

    1990-01-01

    The volumetric and chemical control circuit has an expansion tank with at least one water-steam chamber connected to the primary circuit by a sampling pipe and a reinjection pipe. The sampling pipe feeds jet pumps controlled by valves. An action on these valves and pumps regulates the volume of the water in the primary circuit. A safety pipe controlled by a flap automatically injects water from the chamber into the primary circuit in case of ruptures. The auxiliary circuit has also systems for purifying the water and controlling the boric acid and hydrogen content [fr

  9. Assessment and management of ageing of major nuclear power plant components important to safety. Primary piping in PWRs

    International Nuclear Information System (INIS)

    2003-07-01

    guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. This report addresses the primary piping in PWRs including main coolant piping, surge and spray lines, Class 1 piping in attached systems, and small diameter piping that cannot be isolated from the primary coolant system. Maintaining the structural integrity of this piping throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

  10. Quadriceps Tendon Rupture and Contralateral Patella Tendon Avulsion Post Primary Bilateral Total Knee Arthroplasty: A Case Report

    Directory of Open Access Journals (Sweden)

    Gaurav Sharma

    2016-07-01

    Full Text Available Background: Extensor mechanism failure secondary to knee replacement could be due to tibial tubercle avulsion, Patellar tendon rupture, patellar fracture or quadriceps tendon rupture. An incidence of Patella tendon rupture of 0.17% and Quadriceps tendon rupture of around 0.1% has been reported after Total knee arthroplasty. These are considered a devastating complication that substantially affects the clinical results and are challenging situations to treat with surgery being the mainstay of the treatment. Case Description: We report here an interesting case of a patellar tendon rupture of one knee and Quadriceps tendon rupture of the contralateral knee following simultaneous bilateral knee replacement in a case of inflammatory arthritis patient. End to end repair for Quadriceps tear and augmentation with Autologous Hamstring tendon graft was done for Patella tendon rupture. OUTCOME: Patient was followed up for a period of 1 year and there was no Extension lag with a flexion of 100 degrees in both the knees. DISCUSSION: The key learning points and important aspects of diagnosing these injuries early and the management techniques are described in this unique case of bilateral extensor mechanism disruption following knee replacements.

  11. Ruptured eardrum

    Science.gov (United States)

    ... eardrum ruptures. After the rupture, you may have: Drainage from the ear (drainage may be clear, pus, or bloody) Ear noise/ ... doctor to see the eardrum. Audiology testing can measure how much hearing has been lost. Treatment You ...

  12. Cutting of CO2 primary circuit pipes of G 2/G 3 using explosive charges

    International Nuclear Information System (INIS)

    Imbard, G.; Le Goaller, C.; Ravera, J.P.; Bonnier, Y.; Guilbert, J.P.; Puntous, R.

    1994-01-01

    The objective of this work is to cut large diameter contaminated pipes from the CO 2 cooling system of the gas-cooled reactors by means of explosive charges and to use the resulting shock wave to remove part of the contamination fixed inside the pipe. Two types of tests have been conducted using different explosives in different forms (the decontamination and the cutting tests) and are described. After testing the cutting modules and decontamination fuses, the effects of the detonations on the environment have been measured and were greater than expected. A metal containment device was therefore designed to absorb part of the energy dissipated by the shock wave and retain the debris from the explosions. A description of the tests conducted for this purpose is given. (O.L.). 7 figs., 3 tabs

  13. Compilation of references, data sources and analysis methods for LMFBR primary piping system components

    International Nuclear Information System (INIS)

    Reich, M.; Esztergar, E.P.; Ellison, E.G.; Erdogan, F.; Gray, T.G.F.; Wells, C.W.

    1977-03-01

    A survey and review program for application of fracture mechanics methods in elevated temperature design and safety analysis has been initiated in December of 1976. This is the first of a series of reports, the aim of which is to provide a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction, reliability and safety analysis of piping components in nuclear plants undergoing sub-creep and elevated temperature service conditions

  14. Development of an ISI robot for the fast breeder reactor MONJU primary heat transfer system piping

    International Nuclear Information System (INIS)

    Tagawa, Akihiro; Ueda, Masashi; Yamashita, Takuya; Narisawa, Masataka; Haga, Kouichi

    2011-01-01

    This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55degC; piping surface, 80degC) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB). (author)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Lethal Ultra-Early Subarachnoid Hemorrhage Due to Rupture of De Novo Aneurysm 5 Months After Primary Aneurysmatic Subarachnoid Hemorrhage.

    Science.gov (United States)

    Walter, Johannes; Unterberg, Andreas W; Zweckberger, Klaus

    2018-05-01

    Approximately 1% of all patients surviving rupture of a cerebral aneurysm suffer from a second aneurysmatic subarachnoid hemorrhage later in their lives, 61% of which are caused by rupture of a de novo aneurysm. Latency between bleedings is usually many years, and younger patients tend to achieve better outcomes from a second subarachnoid hemorrhage. We report an unusual case of lethal ultra-early rupture of a de novo aneurysm of the anterior communicating artery only 5 months after the initial subarachnoid hemorrhage and complete coiling in a young, healthy male patient. Despite complete aneurysm obliteration, young age, and good recovery, patients may be subjected to secondary subarachnoid hemorrhages from de novo aneurysms after only a few months of the initial bleeding. Early-control magnetic resonance angiography might hence be advisable. Copyright © 2018 Elsevier Inc. All rights reserved.

  17. Welded joints engineering design of the primary circuit, surge line and main steam piping of the Angra 2 reactor

    International Nuclear Information System (INIS)

    Volta, Angelo Roberto; Couto, Jose Gonzalo Villaverde

    1995-01-01

    The erection of nuclear systems of a Nuclear Power Station is under international requests, that results in a detailed elaboration of documents for the performance of welds. NUCLEN as an engineering design company, responsible for the erection of Angra 2, developed a suitable software program for the elaboration of welding procedure qualifications, tests and examination sequence plans and heat treatment plans applied to primary circuit, surgeline and main steam piping. The paper shows the employed methodology for the elaboration of these documents, as well as the requested engineering design of welding technology and testability in order to assure the stipulated quality level, according to requirements of the specifications, codes and norms. (author). 6 refs

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  19. Spontaneous rupture of adrenal metastasis from hepatocellular carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Chae Hun; Kim, Hyun Jin; Park, Soo Youn; Hwang, Seong Su; Choi, Hyun Joo [St. Vincent Hospital, Suwon (Korea, Republic of)

    2007-03-15

    Rupture of adrenal tumor from various primary origins is a rather rare event. We report here on a ruptured adrenal metastasis from hepatocellular carcinoma, and this ruptured metastasis was observed at the time of the initial diagnosis.

  20. 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.

  1. Rupture disc

    International Nuclear Information System (INIS)

    Newton, R.G.

    1977-01-01

    The intermediate heat transport system for a sodium-cooled fast breeder reactor includes a device for rapidly draining the sodium therefrom should a sodium-water reaction occur within the system. This device includes a rupturable member in a drain line in the system and means for cutting a large opening therein and for positively removing the sheared-out portion from the opening cut in the rupturable member. According to the preferred embodiment of the invention the rupturable member includes a solid head seated in the end of the drain line having a rim extending peripherally therearound, the rim being clamped against the end of the drain line by a clamp ring having an interior shearing edge, the bottom of the rupturable member being convex and extending into the drain line. Means are provided to draw the rupturable member away from the drain line against the shearing edge to clear the drain line for outflow of sodium therethrough

  2. Pipe line construction for reactor containment buildings

    International Nuclear Information System (INIS)

    Aoki, Masataka; Yoshinaga, Toshiaki

    1978-01-01

    Purpose: To prevent the missile phenomenon caused by broken fragments due to pipe whip phenomenon in a portion of pipe lines connected to a reactor containment from prevailing to other portions. Constitution: Various pipe lines connected to the pressure vessel are disposed at the outside of the containments and they are surrounded with a plurality of protection partition walls respectively independent from each other. This can eliminate the effect of missile phenomena upon pipe rupture from prevailing to the pipe lines and instruments. Furthermore this can afford sufficient spaces for the pipe lines, as well as for earthquake-proof supports. (Horiuchi, T.)

  3. Ratchetting behavior of primary heat transport (PHT) piping material SA-333 carbon steel subjected to cyclic loads at room temperature

    International Nuclear Information System (INIS)

    Kulkarni, S.; Desai, Y.M.; Kant, T.; Reddy, G.R.; Gupta, C.; Chakravarthy, J.K.

    2004-01-01

    Ratchetting behavior of SA-333 Gr. 6 carbon steel used as primary heat transport (PHT) piping material has been investigated with three constitutive models proposed by Armstrong-Frederick, Chaboche and Ohno-Wang involving different hardening rules. Performance of the above mentioned models have been evaluated for a broad set of uniaxial and biaxial loading histories. The uniaxial ratchetting simulations have been performed for a range of stress ratios (R) by imposing different stress amplitudes and mean stress conditions. Numerical simulations indicated significant ratchetting and opening of hysteresis loop for negative stress ratio with constant mean stress. Application of cyclic stress without mean stress (R = -1.0) has been observed to produce negligible ratchet-strain accumulation in the material. Simulation under the biaxial stress condition was based on modeling of an internally pressurized thin walled pipe subjected to cyclic bending load. Numerical results have been validated with the experiments as per simulation conditions. All three models have been found to predict the observed accumulation of circumferential strain with increasing number of cycles. However, the Armstrong Frederick (A-F) model was found to be inadequate in simulating the ratchetting response for both uniaxial as well as biaxial loading cases. The A-F model actually over-predicted the ratchetting strain in comparison with the experimental strain values. On the other hand, results obtained with the Chaboche and the Ohno-Wang models for both the uniaxial as well as biaxial loading histories have been observed to closely simulate the experimental results. The Ohno-Wang model resulted in better simulation for the presents sets of experimental results in comparison with the Chaboche model. It can be concluded that the Ohno-Wang model suited well compared to the Chaboche model for above sets of uniaxial and biaxial loading histories. (authors)

  4. Vibration analysis of primary inlet pipe line during steady state and transient conditions of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Ayazuddin, S.K.; Qureshi, A.A.; Hayat, T.

    1997-11-01

    The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from hold-up tank to the reactor pool of Pakistan Research Reactor-1 (PARR-1). The section of the pipeline from heat exchangers to the valve pit is hanger supported in the pump room and the rest of the section from valve pit to the reactor pool is embedded. The PW-IPL is subjected to steady state and transient vibrations. The reactor pumps, which drive the coolant through various circuits mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of the check valve (water hammer). The ASME Boiler and Pressure Vessel code provides data about the acceptable limits of stresses related to the primary static stress due to steady state vibrations. However, due to complexity in the pipe structure, stresses related to the transient vibrations are neglected in the code. In this report attempt has been made to analyzed both steady state and transient vibrations of PW-IPL of PARR-1. Since, both the steady state and transient vibrations affect the hanger-supported section of the PW-IPL, therefore, it was selected for vibration test measurements. In the analysis vibration data was compared with the allowable limits and estimations of maximum pressure build-up, eflection, natural frequency, tensile and shear load on hanger support, and the ratio of maximum combine stress to the allowable load were made. (author)

  5. Globe Rupture

    Directory of Open Access Journals (Sweden)

    Reid Honda

    2017-07-01

    Full Text Available History of present illness: A 46-year-old male presented to the emergency department (ED with severe left eye pain and decreased vision after tripping and striking the left side of his head on the corner of his wooden nightstand. The patient arrived as an inter-facility transfer for a suspected globe rupture with a protective eye covering in place; thus, further physical examination of the eye was not performed by the emergency physician in order to avoid further leakage of aqueous humor. Significant findings: The patient’s computed tomography (CT head demonstrated a deformed left globe, concerning for ruptured globe. The patient had hyperdense material in the posterior segment (see green arrow, consistent with vitreous hemorrhage. CT findings that are consistent with globe rupture may include a collapsed globe, intraocular air, or foreign bodies. Discussion: A globe rupture is a full-thickness defect in the cornea, sclera, or both.1 It is an ophthalmologic emergency. Globe ruptures are almost always secondary to direct perforation via a penetrating mechanism; however, it can occur due to blunt injury if the force generated creates sufficient intraocular pressure to tear the sclera.2 Globes most commonly rupture at the insertions of the intraocular muscles or at the limbus. They are associated with a high rate of concomitant orbital floor fractures.2,3 Possible physical examination findings include a shallow anterior chamber on slit-lamp exam, hyphema, and an irregular “teardrop” pupil. Additionally, a positive Seidel sign, which is performed by instilling fluorescein in the eye and then examining for a dark stream of aqueous humor, is indicative of a globe rupture.4 CT is often used to assess for globe rupture; finds of a foreign body, intraocular air, abnormal contour or volume of the globe, or disruption of the sclera suggest globe rupture.2 The sensitivity of CT scan for diagnosis of globe rupture is only 75%; thus, high clinical

  6. Ruptured Spleen

    Science.gov (United States)

    ... be caused by various underlying problems, such as mononucleosis and other infections, liver disease, and blood cancers. ... cause a ruptured spleen. For instance, people with mononucleosis — a viral infection that can cause an enlarged ...

  7. Review of ASME code criteria for control of primary loads on nuclear piping system branch connections and recommendations for additional development work

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Gwaltney, R.C.; Moore, S.E.

    1993-11-01

    This report collects and uses available data to reexamine the criteria for controlling primary loads in nuclear piping branch connections as expressed in Section III of the ASME Boiler and Pressure Vessel Code. In particular, the primary load stress indices given in NB-3650 and NB-3683 are reexamined. The report concludes that the present usage of the stress indices in the criteria equations should be continued. However, the complex treatment of combined branch and run moments is not supported by available information. Therefore, it is recommended that this combined loading evaluation procedure be replaced for primary loads by the separate leg evaluation procedure specified in NC/ND-3653.3(c) and NC/ND-3653.3(d). No recommendation is made for fatigue or secondary load evaluations for Class 1 piping. Further work should be done on the development of better criteria for treatment of combined branch and run moment effects

  8. Implementation of the structural integrity analysis for PWR primary components and piping

    International Nuclear Information System (INIS)

    Pellissier-Tanon, A.

    1982-01-01

    The trends on the definition, the assessment and the application of fracture strength evaluation methodology, which have arisen through experience in the design, construction and operation of French 900-MW plants are reviewed. The main features of the methodology proposed in a draft of Appendix ZG of the RCC-M code of practice for the design verification of fracture strength of primary components are presented. The research programs are surveyed and discussed from four viewpoints, first implementation of the LEFM analysis, secondly implementation of the fatigue crack propagation analysis, thirdly analysis of vessel integrity during emergency core cooling, and fourthly methodology for tear fracture analysis. (author)

  9. Pipe support

    International Nuclear Information System (INIS)

    Pollono, L.P.

    1979-01-01

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

  10. Oscillating heat pipes

    CERN Document Server

    Ma, Hongbin

    2015-01-01

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

  11. Structural analysis of the as-built IEA-R1 primary coolant piping system using a complete three dimensional model

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Martins, Lucas B.; Marcolin, Gabriel; Mattar Neto, Miguel

    2011-01-01

    IEA-R1 is an open pool type research reactor, moderated by light water and upgraded from 2 MW to 5 MW of operating power level. Heat generated in the reactor core is removed by a coolant system divided in two circuits, primary and secondary, composed by pumps, piping, heat exchangers, cooling tower, and some other auxiliary components. The 5 MW operating power level is now possible due to a modernization program started in 1996. As a part of the modernization program, ageing assessment studies recommend the replacement of one of the two heat exchangers in the circuit. To manage this replacement, modifications in the layout of the primary and secondary piping and supporting systems were performed, based on preliminary stress analysis study. Then, the aim of this work is to present the final stress analysis of the primary circuit. To reach this and taking the modifications of the primary into account, a 3D model of the whole circuit, in the as-built condition, was made. Stress results and discussions are shown. (author)

  12. Simultaneous bilateral patellar tendon rupture ?

    OpenAIRE

    Moura, Diogo Lino; Marques, Jos? Pedro; Lucas, Francisco Manuel; Fonseca, Fernando Pereira

    2016-01-01

    Bilateral patellar tendon rupture is a rare entity, often associated with systemic diseases and patellar tendinopathy. The authors report a rare case of a 34-year-old man with simultaneous bilateral rupture of the patellar tendon caused by minor trauma. The patient is a retired basketball player with no past complaints of chronic knee pain and a history of steroid use. Surgical management consisted in primary end-to-end tendon repair protected temporarily with cerclage wiring, followed by a s...

  13. Stress-assisted, microbial-induced corrosion of stainless steel primary piping and other aging issues at the Omega West Reactor

    International Nuclear Information System (INIS)

    Andrade, A.

    1995-01-01

    After the discovery of cooling system leak of about 284 liters per twenty-four (24) hour period, an investigation determined that the 76.2-cm diameter, 33.5-m long stainless-steel (304) OWR delay line was losing water at the same nominal rate. An excavation effort revealed that a circumferential crack, approximately 0.0025 cm in width, extended around the bottom half of the delay line. In addition, other evidence of what appeared to be microcracking and pitting that originated at random nucleated sites around the pipe were also found. Results of destructive analysis and nondestructive testing allowed Los Alamos staff to conclude that the direct cause for the main crack and other pitting resulted from stress-assisted, microbial-induced corrosion of the stainless steel primary piping. The results also indicated that microbial action from bacteria that are normally present in earth can be extremely harmful to stainless- steel piping under certain conditions. Other potential problems that could have also eventually led to a permanent shutdown of the OWR were discussed. These problems, although never encountered nor associated with the current shutdown, were identified in aging studies and are associated with: (1) the water-cooled, bismuth gamma-ray shield and, (2) the aluminum thermal column head seal that prevents reactor vessel water from entering into the graphite-filled thermal column

  14. 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)

  15. B Plant process piping replacement feasibility study

    International Nuclear Information System (INIS)

    Howden, G.F.

    1996-01-01

    Reports on the feasibility of replacing existing embedded process piping with new more corrosion resistant piping between cells and between cells and a hot pipe trench of a Hanford Site style canyon facility. Provides concepts for replacement piping installation, and use of robotics to replace the use of the canyon crane as the primary means of performing/supporting facility modifications (eg, cell lining, pipe replacement, equipment reinstallation) and operational maintenenace

  16. 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)

  17. Development of an on-site measurement method for residual stress in primary system piping of nuclear power plants

    International Nuclear Information System (INIS)

    Maekawa, Akira; Takahashi, Shigeru; Fujiwara, Masaharu

    2014-01-01

    In residual stress measurement for large-scale pipes and vessels in high radiation areas and highly contaminated areas of nuclear plants, it is difficult to bring the radioactivated pipes and vessels out of the areas as they are. If they can brought out, it is very burdensome to handle them for the measurement. Development of an on-site measurement method of residual stress which can be quickly applied and has sufficient measurement accuracy is desirable. In this study, a new method combining an electric discharge skim-cut method with a microscopic strain measurement method using markers was proposed to realize the on-site residual stress measurement on pipes in high radiation areas and highly contaminated areas. In the electric discharge skim-cut method, a boat-type sample is skimmed out of a pipe outer/inner surface using electric discharge machining and released residual stress is measured. The on-site measurement of residual stress by the method can be done using a small, portable electric discharge machine. In the microscopic strain measurement method using markers, the residual stress is estimated by microscopic measurement of the distance between markers after the stress release. The combination of both methods can evaluate the residual stress with the same accuracy as conventional methods offer and it can achieve reduction of radiation exposure in the measurement because the work is done simply and rapidly. In this study, the applicability of the electric discharge skim-cut method was investigated because the applicability of the microscopic strain measurement method using markers was confirmed previously. The experimental examination clarified the applicable conditions for the residual stress measurement with the same accuracy as the conventional methods. Furthermore, the electric discharge machining conditions using pure water as the machining liquid was found to eliminate the amount of liquid radioactive waste completely. (author)

  18. Hemothorax caused by spontaneous rupture of hepatocellular carcinoma in the pleural cavity: A case report

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Hin Hee; Ohm, Joon Young [Dept. of Radiology, Chungnam National University Hospital, Daejeon (Korea, Republic of); Kim, Song Soo; Kim, Jin Hwan [Dept. of Radiology, Chungnam National University School of Medicine, Daejeon(Korea, Republic of)

    2017-07-15

    Hemothorax resulting from ruptured hepatocellular carcinoma (HCC) is extremely rare and is generally caused by ruptured intrathoracic metastatic lesions. However, we report a rare case of hemothorax resulting from intrathoracic rupture of primary HCC.

  19. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  20. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Bain, R.A.; Testa, M.F.

    1994-01-01

    Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

  1. 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

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

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1981-08-01

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

  3. Leak detection in the primary reactor coolant piping of nuclear power plant by applying beam-microphone technology

    International Nuclear Information System (INIS)

    Kasai, Yoshimitsu; Shimanskiy, Sergey; Naoi, Yosuke; Kanazawa, Junichi

    2004-01-01

    A microphone leak detection method was applied to the inlet piping of the ATR-prototype reactor, Fugen. Statistical analysis results showed that the cross-correlation method provided the effective results for detection of a small leakage. However, such a technique has limited application due to significant distortion of the signals on the reactor site. As one of the alternative methods, the beam-microphone provides necessary spatial selectivity and its performance is less affected by signal distortion. A prototype of the beam-microphone was developed and then tested at the O-arai Engineering Center of the Japan Nuclear Cycle Development Institute (JNC). On-site testing of the beam-microphone was carried out in the inlet piping room of an RBMK reactor of the Leningrad Nuclear Power Plant (LNPP) in Russia. A leak sound imitator was used to simulate the leakage sound under the leakage flow condition of 1-3 gpm (0.23-0.7 m 3 /h). Analysis showed that signal distortion does not seriously affect the performance of this method, and that sound reflection may result in the appearance of ghost sound sources. The test results showed that the influences of sound reflection and background noise were smaller at the high frequencies where the leakage location could be estimated with an angular accuracy of 5deg which is the range of localization accuracy required for the leak detection system. (author)

  4. 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

  5. Heat pipe

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. Spontaneous Achilles tendon rupture in alkaptonuria | Mohammed ...

    African Journals Online (AJOL)

    Spontaneous Achilles tendon ruptures are uncommon. We present a 46-year-old man with spontaneous Achilles tendon rupture due to ochronosis. To our knowledge, this has not been previously reported in Sudan literature. The tendon of the reported patient healed well after debridement and primary repairs.

  7. Pipe clamp effects on thin-walled pipe design

    International Nuclear Information System (INIS)

    Lindquist, M.R.

    1980-01-01

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

  8. Metallurgical analysis of high pressure gas pipelines rupture

    International Nuclear Information System (INIS)

    Hasan, F.; Ahmed, F.

    2007-01-01

    On 6 July 2004, two parallel-running gas pipelines (18-inch and 24-inch diameters), in the main transmission network of SNGPL (a gas company in Pakistan) were ruptured. The ruptures occurred in the early hours of the morning about 8 miles downstream of the compressor station AC-4. The ruptures were indicated by the increased gas flow at the outlet of AC-4 (1), first at about 0648 hours and then again about 20 minutes later. The gas escaping from the ruptured lines had caught fire, and the flames had also 'affected' a third parallel-running pipeline of 30-inch diameter, lying next to the 24-inch line. The metallurgical examination of the two ruptured lines showed that the 24-inch line was ruptured with the help of an explosive device that had been placed on the underside of the pipe. An examination of the 18-inch line showed that this pipe had failed as a result of the heating of the pipe-wall, presumably, by the flame emanating from the 24-inch line. These two observations clearly suggested that the 24-inch line was the first to rupture (by explosives), and the fire following this rupture had heated the 18-inch pipe to a temperature where its yield strength was unable to support the inside gas pressure. The 20 minutes time interval between the two ruptures was obviously the time taken by the 18 inch pipe to be heated upto the level where it started to yield. The 30-inch line lying next to the 24-inch line was affected to the extent that its coating had been burnt-off over a length of about 40-50 feet. However, the pipe did not exhibit any signs of deshaping or deformation what-so-ever. A replica metallographic examination indicated that the microstructure of the pipe was not measurably affected by the heat. It was thus decided not to replace the affected part of the 30-inch pipe, but only to re-coat this affected portion. (author)

  9. 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

  10. Simultaneous bilateral patellar tendon rupture.

    Science.gov (United States)

    Moura, Diogo Lino; Marques, José Pedro; Lucas, Francisco Manuel; Fonseca, Fernando Pereira

    2017-01-01

    Bilateral patellar tendon rupture is a rare entity, often associated with systemic diseases and patellar tendinopathy. The authors report a rare case of a 34-year-old man with simultaneous bilateral rupture of the patellar tendon caused by minor trauma. The patient is a retired basketball player with no past complaints of chronic knee pain and a history of steroid use. Surgical management consisted in primary end-to-end tendon repair protected temporarily with cerclage wiring, followed by a short immobilization period and intensive rehabilitation program. Five months after surgery, the patient was able to fully participate in sport activities.

  11. Simultaneous bilateral patellar tendon rupture

    Directory of Open Access Journals (Sweden)

    Diogo Lino Moura

    Full Text Available ABSTRACT Bilateral patellar tendon rupture is a rare entity, often associated with systemic diseases and patellar tendinopathy. The authors report a rare case of a 34-year-old man with simultaneous bilateral rupture of the patellar tendon caused by minor trauma. The patient is a retired basketball player with no past complaints of chronic knee pain and a history of steroid use. Surgical management consisted in primary end-to-end tendon repair protected temporarily with cerclage wiring, followed by a short immobilization period and intensive rehabilitation program. Five months after surgery, the patient was able to fully participate in sport activities.

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

    International Nuclear Information System (INIS)

    Lydell, B.

    1999-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

  14. 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

  15. Development of nonlinear dynamic analysis program for nuclear piping systems

    International Nuclear Information System (INIS)

    Kamichika, Ryoichi; Izawa, Masahiro; Yamadera, Masao

    1980-01-01

    In the design for nuclear power piping, pipe-whip protection shall be considered in order to keep the function of safety related system even when postulated piping rupture occurs. This guideline was shown in U.S. Regulatory Guide 1.46 for the first time and has been applied in Japanese nuclear power plants. In order to analyze the dynamic behavior followed by pipe rupture, nonlinear analysis is required for the piping system including restraints which play the role of an energy absorber. REAPPS (Rupture Effective Analysis of Piping Systems) has been developed for this purpose. This program can be applied to general piping systems having branches etc. Pre- and post- processors are prepared in this program in order to easily input the data for the piping engineer and show the results optically by use of a graphic display respectively. The piping designer can easily solve many problems in his daily work by use of this program. This paper describes about the theoretical background and functions of this program and shows some examples. (author)

  16. Component external leakage and rupture frequency estimates

    International Nuclear Information System (INIS)

    Eide, S.A.; Khericha, S.T.; Calley, M.B.; Johnson, D.A.; Marteeny, M.L.

    1991-11-01

    In order to perform detailed internal flooding risk analyses of nuclear power plants, external leakage and rupture frequencies are needed for various types of components - piping, valves, pumps, flanges, and others. However, there appears to be no up-to-date, comprehensive source for such frequency estimates. This report attempts to fill that void. Based on a comprehensive search of Licensee Event Reports (LERs) contained in Nuclear Power Experience (NPE), and estimates of component populations and exposure times, component external leakage and rupture frequencies were generated. The remainder of this report covers the specifies of the NPE search for external leakage and rupture events, analysis of the data, a comparison with frequency estimates from other sources, and a discussion of the results

  17. IPM Pipe

    Science.gov (United States)

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

  18. Pipe grabber

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-05-15

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

  19. 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.)

  20. Mechanical properties of roll extruded nuclear reactor piping

    International Nuclear Information System (INIS)

    Steichen, J.M.; Knecht, R.L.

    1975-01-01

    The elevated temperature mechanical properties of large diameter (28 inches) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of Type 316H stainless steel piping material used exhibited consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceeded values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050 0 F for times to 10,000 hours. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900 0 F and that for temperatures of 1050 0 F and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations. (U.S.)

  1. Radiation-resistance of polyurethane pipes for cooling liquid in BES III

    International Nuclear Information System (INIS)

    Li Xunfeng; Zheng Lifang; Ji Quan; Wu Ping; Wang Li

    2009-01-01

    Gamma ray radiation and neutron radiation are predominant in the working conditions of BES III, and the radiation-resistance aging of polyurethane pipes is very important in this condition, as the pipes of cooling liquid for beam pipe and SCQ (superconducting quadrupole) vacuum pipe in BESIII. Polyester polyurethane pipes and polyether polyurethane pipes were irradiated by gamma ray and neutron. The radiation doses were as much as ten years' doses in BES. Pressure test, FTIR and thermal analysis were used to study the radiation-resistance of these two kinds of polyurethane pipes. The results show that the radiation-resistance and thermal stability of polyester polyurethane pipes are prior to those of polyether polyurethane pipes, and the pressure resistance of polyester polyurethane pipes is almost maintained after the irradiation by gamma ray and neutron, but the polyether polyurethane pipes can be aged and ruptured after the irradiation by neutron. (authors)

  2. Triple Achilles Tendon Rupture: Case Report.

    Science.gov (United States)

    Saxena, Amol; Hofer, Deann

    We present a case report with 1-year follow-up data of a 57-year-old male soccer referee who had sustained an acute triple Achilles tendon rupture injury during a game. His triple Achilles tendon rupture consisted of a rupture of the proximal watershed region, a rupture of the main body (mid-watershed area), and an avulsion-type rupture of insertional calcific tendinosis. The patient was treated surgically with primary repair of the tendon, including tenodesis with anchors. Postoperative treatment included non-weightbearing for 4 weeks and protected weightbearing until 10 weeks postoperative, followed by formal physical therapy, which incorporated an "antigravity" treadmill. The patient was able to return to full activity after 26 weeks, including running and refereeing, without limitations. Copyright © 2017 The American College of Foot and Ankle Surgeons. Published by Elsevier Inc. All rights reserved.

  3. 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.

  4. Predicting creep rupture from early strain data

    International Nuclear Information System (INIS)

    Holmstroem, Stefan; Auerkari, Pertti

    2009-01-01

    To extend creep life modelling from classical rupture modelling, a robust and effective parametric strain model has been developed. The model can reproduce with good accuracy all parts of the creep curve, economically utilising the available rupture models. The resulting combined model can also be used to predict rupture from the available strain data, and to further improve the rupture models. The methodology can utilise unfailed specimen data for life assessment at lower stress levels than what is possible from rupture data alone. Master curves for creep strain and rupture have been produced for oxygen-free phosphorus-doped (OFP) copper with a maximum testing time of 51,000 h. Values of time to specific strain at given stress (40-165 MPa) and temperature (125-350 deg. C) were fitted to the models in the strain range of 0.1-38%. With typical inhomogeneous multi-batch creep data, the combined strain and rupture modelling involves the steps of investigation of the data quality, extraction of elastic and creep strain response, rupture modelling, data set balancing and creep strain modelling. Finally, the master curves for strain and rupture are tested and validated for overall fitting efficiency. With the Wilshire equation as the basis for the rupture model, the strain model applies classical parametric principles with an Arrhenius type of thermal activation and a power law type of stress dependence for the strain rate. The strain model also assumes that the processes of primary and secondary creep can be reasonably correlated. The rupture model represents a clear improvement over previous models in the range of the test data. The creep strain information from interrupted and running tests were assessed together with the rupture data investigating the possibility of rupture model improvement towards lower stress levels by inverse utilisation of the combined rupture based strain model. The developed creep strain model together with the improved rupture model is

  5. Rupture politique et enseignement de l'histoire en Afrique du Sud: Les manuels de l'enseignement primaire. (Political rupture and the teaching of history in South Africa: Handbooks of primary-level teaching.)

    Science.gov (United States)

    Carpentier, Claude

    2000-01-01

    Analyzes the content of South African history text books from the primary level upwards and from the 1980s to the most recent publications. The considerable changes in the content of these books seem to be based on two different theoretical models: the multicultural model and the notion of the universality of humanity. (Contains 40 references.)…

  6. Rupture of primigravid uterus and recurrent rupture

    Directory of Open Access Journals (Sweden)

    Nahreen Akhtar

    2016-08-01

    Full Text Available Uterine rupture is a deadly obstetrical emergency endangering the life of both mother and fetus. In Bangladesh, majority of deliveries arc attended by unskilled traditional birth attendant and maternal mortality is still quite high. It is rare Ln developed country but unfortunately it is common in a developing country like Bangladesh. We report a case history of a patient age 32yrs from Daudkandi, Comilla admitted with H/0 previous two rupture uterus and repair with no living issue. We did caesarean section at her 31+ weeks of pregnancy when she developed Jabour pain. A baby of 1.4 kg was delivered. During cesarean section, focal rupture was noted in previous scar of rupture. Unfortunately the baby expired in neonatal ICU after 36 hours.

  7. Effect of ferrite on the precipitation of σ phase in cast austenitic stainless steel used for primary coolant pipes of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yongqiang; Li, Na, E-mail: wangyongqiang1124@163.com [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology, Beijing (China)

    2017-11-15

    The effect of ferrite phase on the precipitation of σ phase in a Z3CN20.09M cast austenitic stainless steel (CASS) used for primary coolant pipes of pressurized water reactor (PWR) nuclear power plants was investigated by using isothermal heat-treatment, optical microscopy (OM), transmission electron microscopy (TEM) and electron probe microanalysis (EPMA) techniques. The influence of different morphologies and volume fractions of ferrite in the σ phase formation mechanism was discussed. The amount of σ phase precipitated in all specimens with different microstructures increased with increasing of aging time, however, the precipitation rate is significant different. The formation of σ phase in specimens with the coarsest ferrite and the dispersively smallest ferrite is slowest. The lowest level Cr content in ferrite and fewest α/γ interfaces in specimen are the main reasons for the slowest σ precipitation due to they are unfavorable for the kinetics and thermodynamics of phase transformation respectively. By contraries, the fastest formation of σ phase takes place in specimens with narrow and long ferrite due to the most α/γ interfaces and higher Cr content in ferrite which are beneficial for preferential nucleation and formation thermodynamics of σ phase. (author)

  8. Pipe failure probability - the Thomas paper revisited

    International Nuclear Information System (INIS)

    Lydell, B.O.Y.

    2000-01-01

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

  9. Hepatic rupture in preeclampsia

    International Nuclear Information System (INIS)

    Winer-Muram, H.T.; Muram, D.; Salazar, J.; Massie, J.D.

    1985-01-01

    The diagnosis of hepatic rupture in patients with pregnancy-induced hypertension (preeclampsia and eclampsia) is rarely made preoperatively. Diagnostic imaging can be utilized in some patients to confirm the preoperative diagnosis. Since hematoma formation precedes hepatic rupture, then, when diagnostic modalities such as sonography and computed tomography identify patients with hematomas, these patients are at risk of rupture, and should be hospitalized until the hematomas resolve

  10. Development, analysis, and evaluation of a commercial software framework for the study of Extremely Low Probability of Rupture (xLPR) events at nuclear power plants.

    Energy Technology Data Exchange (ETDEWEB)

    Kalinich, Donald A.; Helton, Jon Craig; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01

    Sandia National Laboratories (SNL) participated in a Pilot Study to examine the process and requirements to create a software system to assess the extremely low probability of pipe rupture (xLPR) in nuclear power plants. This project was tasked to develop a prototype xLPR model leveraging existing fracture mechanics models and codes coupled with a commercial software framework to determine the framework, model, and architecture requirements appropriate for building a modular-based code. The xLPR pilot study was conducted to demonstrate the feasibility of the proposed developmental process and framework for a probabilistic code to address degradation mechanisms in piping system safety assessments. The pilot study includes a demonstration problem to assess the probability of rupture of DM pressurizer surge nozzle welds degraded by primary water stress-corrosion cracking (PWSCC). The pilot study was designed to define and develop the framework and model; then construct a prototype software system based on the proposed model. The second phase of the project will be a longer term program and code development effort focusing on the generic, primary piping integrity issues (xLPR code). The results and recommendations presented in this report will be used to help the U.S. Nuclear Regulatory Commission (NRC) define the requirements for the longer term program.

  11. 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)

  12. Renal allograft rupture: US diagnosis

    International Nuclear Information System (INIS)

    Maklad, N.F.

    1987-01-01

    The US appearances in seven pathologically and/or surgically proved cases of renal allograft rupture are presented. These include a triangular or amorphous echogenic area in the cortex and medulla in a polar location, an echogenic band or wavy, branching anechoic lines in the hyperechoic region, a subcapsular hematoma, and an extrarenal hematoma in direct continuity with the echogenic area. Duplex Doppler examination in renal allograft rupture shows marked reduction of absence of the diastolic component of the velocity waveform in the arcuate and interlobar arteries, with reduction in amplitude of the systolic wave form. Correlation of the US appearances with gross and microscopic pathologic findings indicates that the echogenic area is due to an intrarenal hematoma, while the echogenic band represents the cortical laceration with adherent blood clots. The US-duplex Doppler examination should be the primary diagnostic modality in this life-threatening condition

  13. Modeling of hot tensile and short-term creep strength for LWR piping materials under severe accident conditions

    International Nuclear Information System (INIS)

    Harada, Y.; Maruyama, Y.; Chino, E.; Shibazaki, H.; Kudo, T.; Hidaka, A.; Hashimoto, K.; Sugimoto, J.

    2000-01-01

    The analytical study on severe accident shows the possibility of the reactor coolant system (RCS) piping failure before reactor pressure vessel failure under the high primary pressure sequence at pressurized water reactors. The establishment of the high-temperature strength model of the realistic RCS piping materials is important in order to predict precisely the accident progression and to evaluate the piping behavior with small uncertainties. Based on material testing, the 0.2% proof stress and the ultimate tensile strength above 800degC were given by the equations of second degree as a function of the reciprocal absolute temperature considering the strength increase due to fine precipitates for the piping materials. The piping materials include type 316 stainless steel, type 316 stainless steel of nuclear grade, CF8M cast duplex stainless steel and STS410 carbon steel. Also the short-term creep rupture time and the minimum creep rate at high-temperature were given by the modified Norton's Law as a function of stress and temperature considering the effect of the precipitation formation and resolution on the creep strength. The present modified Norton's Law gives better results than the conventional Larson-Miller method. Correlating the creep data (the applied stress versus the minimum creep rate) with the tensile data (the 0.2% proof stress or the ultimate tensile strength versus the strain rate), it was found that the dynamic recrystallization significantly occurred at high-temperature. (author)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  15. Alkali Metal Heat Pipe Life Issues

    International Nuclear Information System (INIS)

    Reid, Robert S.

    2004-01-01

    One approach to fission power system design uses alkali metal heat pipes for the core primary heat-transfer system. Heat pipes may also be used as radiator elements or auxiliary thermal control elements. This synopsis characterizes long-life core heat pipes. References are included where information that is more detailed can be found. Specifics shown here are for demonstration purposes and do not necessarily reflect current Nasa Project Prometheus point designs. (author)

  16. 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.)

  17. 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.)

  18. Multi-Canister overpack necessity of the rupture disk

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Multi-Canister Overpack (MCO) rupture disk precludes the MCO from pressurization above the design limit during transport from the K Basins to the Cold Vacuum Drying (CVD) Facility and prior to connection of the CVD process piping. Removal of the rupture disk from the MCO design would: (a) result in unacceptable dose consequences in the event a thermal runaway accident occurred; (b) increase residual risk; and (c) remove a degree of specificity from the dose calculations. The potential cost savings of removing the rupture disk from the MCO design is offset by the cost of design modifications, changes to hazard analyses and safety analyses, and changes to existing documentation. Retaining the rupture disk mitigates the consequences of MCO overpressurization, and considering the overall economic impacts to the SNF Project, is the most cost effective approach

  19. Splenic rupture masquerading ruptured ectopic pregnancy | Kigbu ...

    African Journals Online (AJOL)

    The classical triad of presentation of delayed menses, irregular vaginal bleeding and abdominal pain may not be encountered at all! Overwhelming features of abdominal pain, amenorrhea, pallor, abdominal tenderness, shifting dullness with positive pregnancy test gave a clinical diagnosis of ruptured ectopic pregnancy.

  20. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1982-01-01

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

  1. Rupture and Continuity in the Socio-emotional Framework of Primary School Reading Books in the later Years of the Franco Regime (1959-1975

    Directory of Open Access Journals (Sweden)

    Ana Badanelli Rubio

    2015-07-01

    Full Text Available Reading books were one of the favorite textual genres of the Franco Regime, used as a privileged vehicle for the transmission of knowledge, values and emotions. This article explores the evolution of their characteristics and contents during the late Franco dictatorship (tardofranquismo. During the first and second phases of Franco Regime, textbooks contained heavy doses of ideological indoctrination, both political and religious. The inculcation of emotions and sentiments in school texts has been demonstrated; in the early decades of the regime, the language employed in the writing of textbooks had the specific objective of manipulating and persuading as a means to control and shape the emotional structure of childhood. There is debate as to what extent and in what manner the nature of the texts (the kinds of readings and their underlying messages were modified under the new economic policy and technocracy. The study examines ruptures and continuities that appear in reading books and in their texts, focusing on publishers and authors on the one hand – the messages and their written and iconic languages – and on the emotions that are transmitted through both routes on the other. The economic opening and political turn of the regime’s later years did lead to certain aesthetic and didactic facelifts in the manuals. However, reading books not only changed their appearance; this study brings to light the changes in socio-emotional schemes, changes that took place slowly and gradually. The projected and stimulated emotions began to be directed more at the world of the children and towards a more constructive way reading, moving away from socio-political objectives

  2. The stress analysis evaluation and pipe support layout for pressurizer discharge system

    International Nuclear Information System (INIS)

    Mao Qing; Wang Wei; Zhang Yixiong

    2000-01-01

    The author presents the stress analysis and evaluation of pipe layout and support adjustment process for Qinshan phase II pressurizer discharge system. Using PDL-SYSPIPE INTERFACE software, the characteristic parameters of the system are gained from 3-D CAD engineering design software PDL and outputted as the input date file format of special pipe stress analysis program SYSPIPE. Based on that, SYSPIPE program fast stress analysis function is applied in adjusting pipe layout , support layout and support types. According to RCC-M standard, the pipe stress analysis and evaluation under deadweight, internal pressure, thermal expansion, seismic, pipe rupture and discharge loads are fulfilled

  3. Situation of secondary system piping wearing in overseas nuclear power plants

    International Nuclear Information System (INIS)

    Chiba, Goro

    2005-01-01

    In consideration of secondary system piping rupture accident at Mihama Nuclear Power Station Unit 3 of Kansai Electric Power Company in August 2004, the management system of secondary pipe wall thickness of Japan and foreign countries were investigated. Moreover, the tendency of the secondary piping thinning events on overseas which the Institute of Nuclear Safety System, Inc. (INSS) obtained was analyzed in order to verify the validity of the Japanese management system. Consequently, it was shown that in the U.S., the fault phenomenon of secondary system piping was reported continuously, and there were also many cases of both degradation and penetration of pipe wall. (author)

  4. Premature rupture of membranes

    Science.gov (United States)

    ... gov/ency/patientinstructions/000512.htm Premature rupture of membranes To use the sharing features on this page, ... water that surrounds your baby in the womb. Membranes or layers of tissue hold in this fluid. ...

  5. Ruptured submitral aneurysm

    Directory of Open Access Journals (Sweden)

    V. Shukla

    2016-09-01

    Full Text Available Submitral aneurysm is a rare entity, with around few hundred cases reported till date. Presentation can be varied. We describe here a case of submitral aneurysm in a young male with rupture into the left atrium cavity.

  6. Achilles Tendon Rupture

    Science.gov (United States)

    ... is a strong fibrous cord that connects the muscles in the back of your calf to your heel bone. If you overstretch your Achilles tendon, it can tear (rupture) completely or just partially. If your Achilles ...

  7. Analysis of independent failure assumptions on postulated secondary high energy line ruptures

    International Nuclear Information System (INIS)

    Hollingsworth, S.D.

    1977-01-01

    Postulated ruptures of the main steam piping in pressurized water reactors result in large amounts of steam being removed from the secondary system. Since the energy removal rate could be many times that of nominal design power, there may be a rapid cooldown of the primary coolant system and a positive addition of reactivity to the reactor core. The Westinghouse protection system design concept incorporates features that trip the reactor, isolate the main steamlines and provide for automatic alternate shutdown capability in the form of boric acid solution injection into the primary coolant system. At the most limiting time in life (end of life) the reactivity calculated to be inserted by the cooldown is sufficient to overcome the shutdown margin predicted to be available from control rods with the most reactive rod in the fully withdrawn position. Because the boron injected into the core may be delayed due to system responses, there is potential that the reactor core could return critical and return to power. The extremely adverse radial power distributions caused by the fully withdrawn control rod causes localized high power densities that could lead to reduced heat transfer capability (DNB). Because of the large amount of stored energy in the reactor coolant system at full power, the cooldown and subsequent return to power is more severe when calculated from a shutdown, hot zero power condition. It is shown that the protection system design has large margins to protect against adverse core effects following a steamline rupture

  8. Ruptured cornual pregnancy

    International Nuclear Information System (INIS)

    Hussain, M.; Yasmeen, H.; Noorani, K.

    2003-01-01

    A case of ruptured cornual pregnancy is presented here. The patient presented with history of 30 weeks gestational amenorrhoea and pain in the lower abdomen and epigastrium for the last seven days. Ultrasound revealed a 29 weeks abdominal pregnancy with blood in the pelvic cavity. On laparotomy; there was a ruptured right cornual pregnancy, treated cornual resection and uterine repair. An alive male baby of one kg weight was delivered from the resected cornua of the uterus. (author)

  9. Rupture loop annex ion exchange RLAIX vault deactivation

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.; Harris, D.L., Westinghouse Hanford

    1996-08-01

    This engineering report documents the deactivation, stabilization and final conditions of the Rupture Loop Annex Ion Exchange (RLAIX) Vault located northwest of the 309 Building`s Plutonium Recycle Test Reactor (PRTR). Twelve ion exchange columns, piping debris, and column liquid were removed from the vault, packaged and shipped for disposal. The vault walls and floor were decontaminated, and portions of the vault were painted to fix loose contamination. Process piping and drains were plugged, and the cover blocks and rain cover were installed. Upon closure,the vault was empty, stabilized, isolated.

  10. Effect of prestrain on ductility and toughness in high strength line pipe steels

    Energy Technology Data Exchange (ETDEWEB)

    Shinohara, Y.; Besson, J. [Paristech, Evry (France). Centre des Materiaux, Mines Paris; Madi, Y. [Ecole d' Ingenieurs, Sceaux (France). Ermess EPF; Paristech, Evry (France). Centre des Materiaux, Mines Paris

    2009-07-01

    The anisotropic plasticity, ductility and toughness of an X100 steel pipeline was investigated both before and after a series of prestraining experiments. The aim of the study was to determine the effect of prestraining on ductility and toughness in high strength pipe steels. Results of the study showed that primary void growth and coalescence was dependent on initial plastic anisotropy and not dependent on tensile prestrain. Secondary void nucleation and growth was not influenced by either the initial plastic anisotropy or by prestraining. Scanning electron microscopy (SEM) studies showed that the main damage mechanism was the void growth of primary dimples. Dimples in the prestrained materials were larger than those observed in materials that had not been prestrained. However, the effect on prestrain on dimple size was limited. Results showed both plastic and rupture anisotropies. It was concluded that prestraining induces a decrease in ductility, but has a significant impact on toughness. 4 refs., 2 tabs., 12 figs.

  11. Development of leak-rupture criteria for axially through-wall cracked pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, X.K.; Leis, B.N. [Battelle Memorial Inst., Columbus, OH (United States)

    2009-07-01

    In this study, J-integral based fracture mechanics were used to characterize the fracture behaviour of a pipeline from crack initiation and stable growth to tearing instability. The aim of the study was to develop an effective leak-rupture criterion for pipelines with axial through-wall cracks (TWC). The J driving force for cracked plates and pipes was estimated using a plastic influence function method. The EPRI estimation scheme and R6 method were used to obtain a general estimate of the J-integral of the pipe. The failure pressures of a TWC pipe at fracture initiation and instability were then modelled in order to define a leak-rupture boundary. Fracture initiation and instability analyses for the TWC pipeline were illustrated. Results of the study showed that the method can be used to improve the accuracy of J driving force estimation and the leak-rupture criterion for pipeline applications. 20 refs., 5 figs.

  12. 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)

  13. Fundamentals of piping design

    CERN Document Server

    Smith, Peter

    2013-01-01

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

  14. Calculation of dynamic hydraulic forces in nuclear plant piping systems

    International Nuclear Information System (INIS)

    Choi, D.K.

    1982-01-01

    A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)

  15. 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

  16. Heat pipes in modern heat exchangers

    International Nuclear Information System (INIS)

    Vasiliev, Leonard L.

    2005-01-01

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

  17. Characteristics of DC electrical braking method of the gas circulator to limit the temperature rise at the heat transfer pipes in the HTTR

    International Nuclear Information System (INIS)

    Kawasaki, K.; Saito, K.; Iyoku, T.

    2001-01-01

    In the safety evaluation of a High Temperature Engineering Test Reactor (HTTR), it must be confirmed that the core has no chance to be damaged and the barrier against the FP release is designed properly not to be affecting the influence of radiation around the reactor site. Especially the maximum temperature of the reactor pressure boundary such as the heat transfer pipes of pressurized water cooler (PWC) must not exceed the permissible values under an anticipated accident such as pipe of rupture in PWC. A requirement for the gas circulator which circulates helium gas in the primary cooling line and the secondary cooling line, is to be braked within 10 seconds by an electrical braking method after the HTTR reactor has scrammed under the accident in PWC. The reason is that the temperature rise of the heat transfer pipe at PWC has to be suppressed when the gas circulator has stopped, the revolution of the gas circulator decreases like the free coast down so that it takes about 90 seconds to be zero and the temperature rise of the pipe in the PWC exceeds the permissible value. By braking within 10 secs., the temperature of the pipe in the PWC reaches about 368 deg. C, less than the permissible value. Using a simplified equivalent circuit of an induction motor, braking time analysis was performed with obtained electrical resistance and inductance. The obtained braking time is about 10 secs., showing close agreement with analysis values. (author)

  18. Failure rate of piping in hydrogen sulphide systems

    International Nuclear Information System (INIS)

    Hare, M.G.

    1993-08-01

    The objective of this study is to provide information about piping failures in hydrogen sulphide service that could be used to establish failures rates for piping in 'sour service'. Information obtained from the open literature, various petrochemical industries and the Bruce Heavy Water Plant (BHWP) was used to quantify the failure analysis data. On the basis of this background information, conclusions from the study and recommendations for measures that could reduce the frequency of failures for piping systems at heavy water plants are presented. In general, BHWP staff should continue carrying out their present integrity and leak detection programmes. The failure rate used in the safety studies for the BHWP appears to be based on the rupture statistics for pipelines carrying sweet natural gas. The failure rate should be based on the rupture rate for sour gas lines, adjusted for the unique conditions at Bruce

  19. Investigation of the cold process pipe rupture mechanism

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    Cryogenic process pipelines are part of the basic subsystem used in installations for fundamental research in physics, as well as in industrial plants which use LNG or liquid nitrogen. The significant increase in importance of cryogenics entails the need to explore phenomena which have direct impact in the design process of cryogenic systems and their safety systems. These aspects are of high priority due to high investment costs, mainly because of safety issues and reliability. One of the issues which requires thorough investigation is the fracture mechanics of gas pipelines in cryogenic conditions. For this subject, importance is placed not only in when the cracks begin to appear, but also in how they form and how quickly they propagate. Currently, there is a lack of reliable research in the available literature in this area. This is often raised as a significant problem for designers, because knowledge in this topic should be reflected e.g., in the sizing calculations of safety valves for the vacuum syste...

  20. 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

  1. Thinned pipe management program of Korean NPPs

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. HPFRCC - Extruded Pipes

    DEFF Research Database (Denmark)

    Stang, Henrik; Pedersen, Carsten

    1996-01-01

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

  3. Pipe drafting and design

    CERN Document Server

    Parisher, Roy A; Parisher

    2000-01-01

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

  4. Experimental Investigation on Corrosion of Cast Iron Pipes

    Directory of Open Access Journals (Sweden)

    H. Mohebbi

    2011-01-01

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

  5. Rupture, waves and earthquakes.

    Science.gov (United States)

    Uenishi, Koji

    2017-01-01

    Normally, an earthquake is considered as a phenomenon of wave energy radiation by rupture (fracture) of solid Earth. However, the physics of dynamic process around seismic sources, which may play a crucial role in the occurrence of earthquakes and generation of strong waves, has not been fully understood yet. Instead, much of former investigation in seismology evaluated earthquake characteristics in terms of kinematics that does not directly treat such dynamic aspects and usually excludes the influence of high-frequency wave components over 1 Hz. There are countless valuable research outcomes obtained through this kinematics-based approach, but "extraordinary" phenomena that are difficult to be explained by this conventional description have been found, for instance, on the occasion of the 1995 Hyogo-ken Nanbu, Japan, earthquake, and more detailed study on rupture and wave dynamics, namely, possible mechanical characteristics of (1) rupture development around seismic sources, (2) earthquake-induced structural failures and (3) wave interaction that connects rupture (1) and failures (2), would be indispensable.

  6. 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)

  7. 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

  8. 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.

  9. OPDE-The international pipe failure data exchange project

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, Bengt [OPDE Clearinghouse, 16917 S. Orchid Flower Trail, Vail, AZ 85641-2701 (United States)], E-mail: boylydell@msn.com; Riznic, Jovica [Canadian Nuclear Safety Commission, Operational Engineering Assessment Division, PO Box 1046, Station B, Ottawa, Ont. K1P 5S9 (Canada)], E-mail: jovica.riznic@cnsc-ccsn.gc.ca

    2008-08-15

    Certain member countries of the Organization for Economic Cooperation and development (OECD) in 2002 established the OECD pipe failure data exchange project (OPDE) to produce an international database on the piping service experience applicable to commercial nuclear power plants. OPDE is operated under the umbrella of the OECD Nuclear Energy Agency (NEA). The Project collects pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large through-wall flow rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of design code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. Currently the OPDE database includes approximately 3,700 records on pipe failure affecting ASME Code Class 1 through 3 and non-safety-related (non-Code) piping. This paper presents the motivations and objectives behind the establishment of the OPDE project. The paper also summarizes the unique data quality considerations that are associated with the reporting and recording of piping component degradation and failure. An overview of the database content is included to place it in perspective relative to past efforts to systematically collect and evaluate service experience data on piping performance. Finally, a brief summary is given of current database application studies.

  10. OPDE-The international pipe failure data exchange project

    International Nuclear Information System (INIS)

    Lydell, Bengt; Riznic, Jovica

    2008-01-01

    Certain member countries of the Organization for Economic Cooperation and development (OECD) in 2002 established the OECD pipe failure data exchange project (OPDE) to produce an international database on the piping service experience applicable to commercial nuclear power plants. OPDE is operated under the umbrella of the OECD Nuclear Energy Agency (NEA). The Project collects pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large through-wall flow rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of design code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. Currently the OPDE database includes approximately 3,700 records on pipe failure affecting ASME Code Class 1 through 3 and non-safety-related (non-Code) piping. This paper presents the motivations and objectives behind the establishment of the OPDE project. The paper also summarizes the unique data quality considerations that are associated with the reporting and recording of piping component degradation and failure. An overview of the database content is included to place it in perspective relative to past efforts to systematically collect and evaluate service experience data on piping performance. Finally, a brief summary is given of current database application studies

  11. 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

  12. Solar heating pipe

    Energy Technology Data Exchange (ETDEWEB)

    Hinson-Rider, G.

    1977-10-04

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

  13. Double rupture disc experience

    International Nuclear Information System (INIS)

    1979-01-01

    Result of these observations, comparisons and evaluations can be summarized in the following list of concerns regarding the use of double rupture discs coupled to the liquid space of a steam generator that is subjected to a large leak sodium water reaction event. Single rupture disc show delayed collapse characteristics in LLTR Series I and double disc assemblies are presumed to be more complex with additional delay before opening to give pressure relief. Delayed failure increases pressures in the IHTS and must be adequately covered by design requirements. With CRBR design, the first disc may fail only partially reducing the loading on the second disc with the result that relief performance may not meet requirements

  14. Ruptured hepatoblastoma treated with primary surgical resection

    African Journals Online (AJOL)

    The remaining part of the left lobe of the liver, including segment IV ... lung, and brain metastasis, and was treated with comfort measures only ... tumor was noted on the right side of the liter. ... invasive and can control the bleeding to allow for a complete workup and ... Our two cases presented here add to the growing body.

  15. Investigation and evaluation of the cause about the loss of split pin which was set to pipe support system of the primary cooling system of Monju

    International Nuclear Information System (INIS)

    Ichikawa, Shoichi; Kawanago, Sho; Nishio, Ryuichi; Wakimoto, Fumitsugu; Fujimura, Tomofumi; Kobayashi, Takanori; Sakamoto, Tsutomu

    2015-07-01

    The loss of the retaining split pins (four pieces) for clevis pin was confirmed at the inspection of the pipe supports in the Monju prototype fast-breeder reactor (hereinafter referred to Monju) in May, 2014. The split pins (two pieces) of ROD RESTRAINT and CONSTANT HANGER were fallen off. The split pins (two pieces) of MECHANICAL SNUBBER were broken at both ends of them. As a result of investigation, a dimple pattern was observed in a fracture surface of broken split pin. This observation result showed that fracture morphology is ductile fracture. A reproduction test, whether split pin was broken by loading the external force to the clevis pin, also gave the same fracture morphology. As the result of all cause investigation, the reason of the broken split pins is that the split pins were loaded shearing stress by the external force loaded to the clevis pin axial direction. The result of the cause investigation and a recurrence prevention measure of this trouble was reported. (author)

  16. Drill pipe bridge plug

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. Structural integrity evaluation of FTL in-pool piping

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y

    1998-05-01

    HANARO fuel test loop will be equipped in HANARO to obtain the development betterment of advanced fuel and materials through the irradiation test. The object of this study is to evaluate the structural integrity of FTL in-pool piping by investigating a dynamic analysis of the loop containing a postulated rupture section. The method to perform the dynamic analysis and structural integrity evaluation caused by the pipe whip in water environment can be a reference for a similar structural integrity evaluation. (author). 7 refs., 39 tabs., 34 figs.

  18. A elastic-plastic model for pipe whip

    International Nuclear Information System (INIS)

    Maneschy, J.E.A.

    1980-04-01

    The dynamic behavior of a cantilever beam simulating a pipe after full rupture at a given cross-section is investigated. This problem, known as pipe whip, has to be analysed within the frame of plastic deformations. The physical model is represented by a cantilever, subjected to a step-load at the free end, and a support designed to absorb the maximum possible kinetic energy of the tube generated by suddenly applied force. The analysis is performed using the Bernoulli theory for straight beams, assuming for the moment-curvature relation a bi-linear law. (author)

  19. 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

  20. Estimative of core damage frequency in IPEN'S IEA-R1 research reactor due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami; Sabundjian, Gaiane; Cabral, Eduardo Lobo Lustosa

    2009-01-01

    The National Commission of Nuclear Energy (CNEN), which is the Brazilian nuclear regulatory commission, imposes safety and licensing standards in order to ensure that the nuclear power plants operate in a safe way. For licensing a nuclear reactor one of the demands of CNEN is the simulation of some accidents and thermalhydraulic transients considered as design base to verify the integrity of the plant when submitted to adverse conditions. The accidents that must be simulated are those that present large probability to occur or those that can cause more serious consequences. According to the FSAR (Final Safety Analysis Report) the initiating event that can cause the largest damage in the core, of the IEA-R1 research reactor at IPEN-CNEN/SP, is the LOCA (Loss of Coolant Accident). The objective of this paper is estimate the frequency of the IEA-R1 core damage, caused by this initiating event. In this paper we analyze the accident evolution and performance of the systems which should mitigate this event: the Emergency Coolant Core System (ECCS) and the isolated pool system. They will be analyzed by means of the event tree. In this work the reliability of these systems are also quantified using the fault tree. (author)

  1. Quality control of stainless steel pipings for nuclear power generation

    International Nuclear Information System (INIS)

    Miki, Minoru; Kitamura, Ichiro; Ito, Hisao; Sasaki, Ryoichi

    1979-01-01

    The proportion of nuclear power in total power generation is increasing recently in order to avoid the concentrated dependence on petroleum resources, consequently the reliability of operation of nuclear power plants has become important. In order to improve the reliability of plants, the reliability of each machine or equipment must be improved, and for the purpose, the quality control at the time of manufacture is the important factor. The piping systems for BWRs are mostly made of carbon steel, and stainless steel pipings are used for the recirculation system cooling reactors and instrumentation system. Recently, grain boundary type stress corrosion cracking has occurred in the heat-affected zones of welded stainless steel pipings in some BWR plants. In this paper, the quality control of stainless steel pipings is described from the standpoint of preventing stress corrosion cracking in BWR plants. The pipings for nuclear power plants must have sufficient toughness so that the sudden rupture never occurs, and also sufficient corrosion resistance so that corrosion products do not raise the radioactivity level in reactors. The stress corrosion cracking occurred in SUS 304 pipings, the factors affecting the quality of stainless steel pipings, the working method which improves the corrosion resistance and welding control are explained. (Kako, I.)

  2. [Achilles tendon rupture].

    Science.gov (United States)

    Thermann, H; Hüfner, T; Tscherne, H

    2000-03-01

    The treatment of acute of Achilles tendon rupture experienced a dynamic development in the last ten years. Decisive for this development was the application of MRI and above all the ultrasonography in the diagnostics of the pathological changes and injuries of tendons. The question of rupture morphology as well as different courses of healing could be now evaluated objectively. These advances led consequently to new modalities in treatment concepts and rehabilitation protocols. The decisive input for improvements of the outcome results and particularly the shortening of the rehabilitation period came with introduction of the early functional treatment in contrast to immobilizing plaster treatment. In a prospective randomized study (1987-1989) at the Trauma Dept. of the Hannover Medical School could show no statistical differences comparing functional non-operative with functional operative therapy with a special therapy boot (Variostabil/Adidas). The crucial criteria for therapy selection results from the sonographically measured position of the tendon stumps in plantar flexion (20 degrees). With complete adaptation of the tendons' ends surgical treatment does not achieve better results than non-operative functional treatment in term of tendon healing and functional outcome. Regarding the current therapeutic standards each method has is advantages and disadvantages. Both, the operative and non-operative functional treatment enable a stable tendon healing with a low risk of re-rupture (1-2%). Meanwhile there is consensus for early functional after-treatment of the operated Achilles' tendons. There seems to be a trend towards non-operative functional treatment in cases of adequate sonographical findings, or to minimal invasive surgical techniques.

  3. Slow rupture of frictional interfaces

    OpenAIRE

    Sinai, Yohai Bar; Brener, Efim A.; Bouchbinder, Eran

    2011-01-01

    The failure of frictional interfaces and the spatiotemporal structures that accompany it are central to a wide range of geophysical, physical and engineering systems. Recent geophysical and laboratory observations indicated that interfacial failure can be mediated by slow slip rupture phenomena which are distinct from ordinary, earthquake-like, fast rupture. These discoveries have influenced the way we think about frictional motion, yet the nature and properties of slow rupture are not comple...

  4. 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

  5. Slow rupture of frictional interfaces

    Science.gov (United States)

    Bar Sinai, Yohai; Brener, Efim A.; Bouchbinder, Eran

    2012-02-01

    The failure of frictional interfaces and the spatiotemporal structures that accompany it are central to a wide range of geophysical, physical and engineering systems. Recent geophysical and laboratory observations indicated that interfacial failure can be mediated by slow slip rupture phenomena which are distinct from ordinary, earthquake-like, fast rupture. These discoveries have influenced the way we think about frictional motion, yet the nature and properties of slow rupture are not completely understood. We show that slow rupture is an intrinsic and robust property of simple non-monotonic rate-and-state friction laws. It is associated with a new velocity scale cmin, determined by the friction law, below which steady state rupture cannot propagate. We further show that rupture can occur in a continuum of states, spanning a wide range of velocities from cmin to elastic wave-speeds, and predict different properties for slow rupture and ordinary fast rupture. Our results are qualitatively consistent with recent high-resolution laboratory experiments and may provide a theoretical framework for understanding slow rupture phenomena along frictional interfaces.

  6. Miniature Heat Pipes

    Science.gov (United States)

    1997-01-01

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

  7. 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.

  8. Rotura prematura de membranas, aspectos de interés para la atención primaria de salud Premature membrane rupture, interesting features for a primary health care

    Directory of Open Access Journals (Sweden)

    Miguel Lugones Botell

    2010-12-01

    Full Text Available Se realizó una revisión sobre aspectos de la rotura prematura de las membranas ovulares, la cual resulta de gran importancia por constituir un tema de mucho interés y tener una morbimortalidad importante en el desarrollo normal del embarazo en cualquiera de sus etapas, sobre todo, antes del término de este. Se destacan aspectos sobre la constitución de las membranas, la clasificación, así como los fundamentos fisiopatológicos y su etiología, enfatizando en la importancia que tienen las infecciones cervicovaginales, y, consecuentemente, en la necesidad de una prevención adecuada en ese sentido, sin prescindir de otros aspectos que también tienen su influencia, como son, las relaciones sexuales y otras complicaciones propias del embarazo (gemelaridad, sangrados de la gestación, presentaciones viciosas, polihidramnios, tabaquismo, etc.. Se resaltan también las complicaciones que pueden ocurrir para el feto y para la madre; y se enfatiza en la necesidad de una adecuada prevención, la cual puede lograrse, fundamentalmente, con el correcto seguimiento de la gestante en el nivel primario de salud.A review on the features related to premature rupture ovular membranes was carried out, which is essential because it is a interesting subject and with a high mortality and morbidity rate in the development of a normal pregnancy in any of its stages, mainly before its term. There are significant features on the membrane's composition, its classification, as well as the physiopathologic basis and its etiology, emphasizing on the significance of cervicovaginal infections and thus in the need of a appropriate prevention in that sense, without disregard of other features that also have its influence including the intercourses and other complications own of pregnancy (twinning, pregnancy bleeding, defective presentations, polyhydramnios, smoking, etc. Also are emphasized the complications that may to occur for fetus and for the mother as well as the

  9. 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.

  10. Riser pipe elevator

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W.; Jimenez, A.F.

    1987-09-08

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

  11. Piping engineering and operation

    International Nuclear Information System (INIS)

    1993-01-01

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

  12. Piping equipment; Materiel petrole

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  13. Safety catching device for pipes in missile shielding cylinders of nuclear power plants

    International Nuclear Information System (INIS)

    Hering, S.; Doll, B.

    1976-01-01

    The safety catching device consists of a steel wire passed in U-shape around the pipe to be caught and supported by two anchor ties embedded in the concrete of the missile shielding cylinder. This flexible catching device is to cause the energy released in case of a pipe rupture to be absorbed and no dangerous bending shesses to be transferred to the walls of the missile shielding cylinder. (UWI) [de

  14. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  15. Untreated silicone breast implant rupture

    DEFF Research Database (Denmark)

    Hölmich, Lisbet R; Vejborg, Ilse M; Conrad, Carsten

    2004-01-01

    Implant rupture is a well-known complication of breast implant surgery that can pass unnoticed by both patient and physician. To date, no prospective study has addressed the possible health implications of silicone breast implant rupture. The aim of the present study was to evaluate whether untre...

  16. U.S. Nuclear Regulatory Commission Extremely Low Probability of Rupture pilot study: xLPR framework model user's guide

    International Nuclear Information System (INIS)

    Kalinich, Donald A.; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-01-01

    For the U.S. Nuclear Regulatory Commission (NRC) Extremely Low Probability of Rupture (xLPR) pilot study, Sandia National Laboratories (SNL) was tasked to develop and evaluate a probabilistic framework using a commercial software package for Version 1.0 of the xLPR Code. Version 1.0 of the xLPR code is focused assessing the probability of rupture due to primary water stress corrosion cracking in dissimilar metal welds in pressurizer surge nozzles. Future versions of this framework will expand the capabilities to other cracking mechanisms, and other piping systems for both pressurized water reactors and boiling water reactors. The goal of the pilot study project is to plan the xLPR framework transition from Version 1.0 to Version 2.0; hence the initial Version 1.0 framework and code development will be used to define the requirements for Version 2.0. The software documented in this report has been developed and tested solely for this purpose. This framework and demonstration problem will be used to evaluate the commercial software's capabilities and applicability for use in creating the final version of the xLPR framework. This report details the design, system requirements, and the steps necessary to use the commercial-code based xLPR framework developed by SNL.

  17. 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)

  18. Blunt traumatic diaphragmatic rupture

    Directory of Open Access Journals (Sweden)

    Antonio Carlos Nogueira

    2011-09-01

    Full Text Available Traumatic injury of the diaphragm ranges from 0.6 to 1.2% and rise up to 5%among patients who were victims of blunt trauma and underwent laparotomy.Clinical suspicion associated with radiological assessment contributes to earlydiagnosis. Isolated diaphragmatic injury has a good prognosis. Generallyworse outcomes are associated with other trauma injuries. Bilateral andright diaphragmatic lesions have worse prognosis. Multi detector computed tomography (MDCT scan of the chest and abdomen provides better diagnosticaccuracy using the possibility of image multiplanar reconstruction. Surgicalrepair via laparotomy and/ or thoracotomy in the acute phase of the injury hasa better outcome and avoids chronic complications of diaphragmatic hernia.The authors present the case of a young male patient, victim of blunt abdominaltrauma due to motor vehicle accident with rupture of the diaphragm, spleenand kidney injuries. The diagnosis was made by computed tomography of thethorax and abdomen and was confirmed during laparotomy.

  19. Costs reduced by innovative plastic distribution pipe use

    International Nuclear Information System (INIS)

    Maxwell, F.W.

    1995-01-01

    As part of a strategic corporate cost-reduction initiative, Pacific Gas and Electric Company's Gas Distribution Group has achieved some quick but significant cash savings. System design, construction, and the purchasing function were areas that produced some fast paybacks while maintaining reliability and safety. The primary savings were made by optimizing pipe specifications to match system operating parameters. This allowed the use of smaller diameter pipes and/or thinner wall pipes which conserved the materials cost of the pipeline. Other realized savings in the form of coiled pipe, purchasing changes, and backfilling specifications are also described

  20. Transients in pipes

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  1. 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…

  2. Piping research program plan

    International Nuclear Information System (INIS)

    1988-09-01

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

  3. Noncondensable gas accumulation phenomena in nuclear power plant piping

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Aoki, Kazuyoshi; Sato, Teruaki; Shida, Akira; Ichikawa, Nagayoshi; Nishikawa, Akira; Inagaki, Tetsuhiko

    2011-01-01

    In the case of the boiling water reactor, hydrogen and oxygen slightly exist in the main steam, because these noncondensable gases are generated by the radiolytic decomposition of the reactor water. BWR plants have taken measures to prevent noncondensable gas accumulation. However, in 2001, the detonation of noncondensable gases occurred at Hamaoka-1 and Brunsbuttel, resulting in ruptured piping. The accumulation phenomena of noncondensable gases in BWR closed piping must be investigated and understood in order to prevent similar events from occurring in the future. Therefore, an experimental study on noncondensable gas accumulation was carried out. The piping geometries for testing were classified and modeled after the piping of actual BWR plants. The test results showed that 1) noncondensable gases accumulate in vertical piping, 2) it is hard for noncondensable gases to accumulate in horizontal piping, and 3) noncondensable gases accumulate under low-pressure conditions. A simple accumulation analysis method was proposed. To evaluate noncondensable gas accumulation phenomena, the three component gases were treated as a mixture. It was assumed that the condensation amount of the vapor is small, because the piping is certainly wrapped with heat insulation material. Moreover, local thermal equilibrium was assumed. This analysis method was verified using the noncondensable gas accumulation test data on branch piping with a closed top. Moreover, an experimental study on drain trap piping was carried out. The test results showed that the noncondensable gases dissolved in the drain water were discharged from the drain trap, and Henry's law could be applied to evaluate the amount of dissolved noncondensable gases in the drain water. (author)

  4. 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.)

  5. Ruptured ectopic pregnancy diagnosed with computed tomography

    International Nuclear Information System (INIS)

    Michalak, Maciej; Żurada, Anna; Biernacki, Maciej; Zygmunt, Kozielec

    2010-01-01

    The rupture of ectopic pregnancy (EP) still remains the primary and direct cause of death in the first trimester of pregnancy. Ultrasonography is known to be a modality of choice in EP diagnostics. We found a severe discrepancy between the frequency of ectopic pregnancies (EP) and the number of available computed tomography (CT) examinations. A 29-year-old woman was admitted to the emergency department with a history of abdominal pain, nausea, vomiting and collapse. Sonographic findings of a suspected EP were unclear. Moreover, not all features of intrauterine pregnancy were present. Due to the patient’s life-threatening condition, an emergency multi-slice CT with MPR and VRT reconstructions was performed, revealing symptoms of a ruptured EP. In the right adnexal area, a well-vascularized, solid-cystic abnormal mass lesion was found. Intraperitoneal hemorrhage was confirmed intraoperatively, and the right fallopian tube with a tubal EP was resected. In the surgery in situ, as well as in the pathological examination of the tumor mass, a human embryo of approximately 1.5 cm in length (beginning of the 8 th week of gestation) was found. Although ultrasonography still remains the first-line imaging examination in EP diagnostics, sometimes the findings of suspected EPs are unclear and not sufficient. The rupture of EP, with serious bleeding and symptoms of shock, may require an emergent pelvic and abdominal CT inspection. A clear correlation was found between the macroscopic CT images and the intraoperatively sampled material

  6. Failure and factors of safety in piping system design

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1993-01-01

    An important body of test and performance data on the behavior of piping systems has led to an ongoing reassessment of the code stress allowables and their safety margin. The codes stress allowables, and their factors of safety, are developed from limits on the incipient yield (for ductile materials), or incipient rupture (for brittle materials), of a test specimen loaded in simple tension. In this paper, we examine the failure theories introduced in the B31 and ASME III codes for piping and their inherent approximations compared to textbook failure theories. We summarize the evolution of factors of safety in ASME and B31 and point out that, for piping systems, it is appropriate to reconsider the concept and definition of factors of safety

  7. A checking device for pipes in which a high pressure fluid is circulated

    International Nuclear Information System (INIS)

    Bauerle, R.D.; Pitt, W.A.; White, M.A.

    1974-01-01

    A checking device for restricting the movements of a pipe in which a high pressure fluid is circulated, should said pipe happen to be ruptured. That device comprises a U-shaped checking, or retaining bar surrounding the pipe, and slightly spaced therefrom at each end of said bar a support member fixed to a frame member of the steam generator and an articulated connection between each of said ends and its respective support-member. That device can be applied to nuclear steam boilers [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-11-01

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

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

    International Nuclear Information System (INIS)

    1984-11-01

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

  10. Pipe drafting and design

    CERN Document Server

    Parisher, Roy A

    2011-01-01

    Pipe Drafting and Design, Third Edition provides step-by-step instructions to walk pipe designers, drafters, and students through the creation of piping arrangement and isometric drawings. It includes instructions for the proper drawing of symbols for fittings, flanges, valves, and mechanical equipment. More than 350 illustrations and photographs provide examples and visual instructions. A unique feature is the systematic arrangement of drawings that begins with the layout of the structural foundations of a facility and continues through to the development of a 3-D model. Advanced chapters

  11. Structural and piping issues in the design certification of advanced reactors

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Development of seamless forged pipe and fitting for BWR recirculation loop piping with improved resistance to intergranular stress corrosion cracking

    International Nuclear Information System (INIS)

    Ohnishi, Keizo; Tsukada, Hisashi; Kobayashi, Masayoshi; Iwadate, Tadao; Ono, Shinichi

    1981-01-01

    As a primary remedy for IGSCC of primary loop piping, especially Recirculation Loop Piping of BWR, extra low carbon stainless steel with high nitrogen content has become to be used. While, in order to make In-service Inspection easier and complete, new design of piping which decrease both number and total length of weld line has been considered. Japan Steel Works has developed the research on large size seamless forged pipe and fitting made from high nitrogen extra low carbon 316 stainless steel. This paper describes the key points of manufacturing technology as well as the material properties, especially strength and intergranular-corrosion and intergranular- stress-corrosion-cracking-resistivities of these forged pipe and fitting. (author)

  13. The diagnosis of breast implant rupture

    DEFF Research Database (Denmark)

    Hölmich, Lisbet R; Vejborg, Ilse; Conrad, Carsten

    2005-01-01

    participated in either one or two study MRI examinations, aiming at determining the prevalence and incidence of silent implant rupture, respectively, and who subsequently underwent explantation. Implant rupture status was determined by four independent readers and a consensus diagnosis of either rupture...... were in fact ruptured at surgery. Thirty-four of the 43 intact implants were described as intact at surgery. When categorising possible ruptures as ruptures, there were one false positive and nine false negative rupture diagnoses at MRI yielding an accuracy of 92%, a sensitivity of 89...

  14. 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.

  15. Simplified pipe gun

    International Nuclear Information System (INIS)

    Sorensen, H.; Nordskov, A.; Sass, B.; Visler, T.

    1987-01-01

    A simplified version of a deuterium pellet gun based on the pipe gun principle is described. The pipe gun is made from a continuous tube of stainless steel and gas is fed in from the muzzle end only. It is indicated that the pellet length is determined by the temperature gradient along the barrel right outside the freezing cell. Velocities of around 1000 m/s with a scatter of +- 2% are obtained with a propellant gas pressure of 40 bar

  16. Stuck pipe prediction

    KAUST Repository

    Alzahrani, Majed

    2016-03-10

    Disclosed are various embodiments for a prediction application to predict a stuck pipe. A linear regression model is generated from hook load readings at corresponding bit depths. A current hook load reading at a current bit depth is compared with a normal hook load reading from the linear regression model. A current hook load greater than a normal hook load for a given bit depth indicates the likelihood of a stuck pipe.

  17. Stuck pipe prediction

    KAUST Repository

    Alzahrani, Majed; Alsolami, Fawaz; Chikalov, Igor; Algharbi, Salem; Aboudi, Faisal; Khudiri, Musab

    2016-01-01

    Disclosed are various embodiments for a prediction application to predict a stuck pipe. A linear regression model is generated from hook load readings at corresponding bit depths. A current hook load reading at a current bit depth is compared with a normal hook load reading from the linear regression model. A current hook load greater than a normal hook load for a given bit depth indicates the likelihood of a stuck pipe.

  18. 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

  19. 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.

  20. Heat pipe dynamic behavior

    Science.gov (United States)

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

    1988-01-01

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

  1. 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.)

  2. Ruptured Ectopic Pregnancy

    Directory of Open Access Journals (Sweden)

    Valentina Park

    2016-09-01

    Full Text Available History of present illness: A 21-year-old female presented with sudden onset suprapubic abdominal pain associated with dysuria. The patient also experienced near syncope during bowel movements three times three days ago without falling or losing consciousness. She denied fever, nausea, and vomiting. She stated that she was five weeks pregnant by last menstrual period. She had an ultrasound a few weeks before that showed no intrauterine pregnancy, but she had not followed up for additional testing. Significant findings: The patient’s serum beta-hCG was 5,637 mIU/mL. The transvaginal ultrasound showed an empty uterus with free fluid posteriorly in the pelvis and Pouch of Douglas (00:00. A 4.5 cm heterogeneous mass was visible in the left adnexa concerning for an ectopic pregnancy (00:10. Discussion: Ectopic pregnancies are a leading cause of maternal morbidity and mortality, as well as decreased fertility.1,2 Differentiating between an ectopic pregnancy and a normal early pregnancy may be difficult, since ultrasound and quantitative beta-hCG may show inconclusive results.3,4 Patients who have used fertility treatment may further complicate the picture because they are at risk for heterotypic pregnancies.5 Ectopic pregnancies most commonly implant in the fallopian tube, but may alternatively implant in the ovary, cervix, abdomen, or uterine cornua.4 Ultrasonography may show an empty uterus, adnexal mass, pelvic free fluid, or an extra-uterine gestational sac, yolk sac, and/or embryo.6 Treatment options for ectopic pregnancy include surgery or methotrexate.2,4 Some patients may be candidates for close outpatient surveillance if the diagnosis is unclear or in very limited cases for early, non-ruptured ectopic pregnancies.2,4

  3. Inelastic analysis of SNR-300 piping

    Energy Technology Data Exchange (ETDEWEB)

    Huebel, H [INTERATOM, Bergisch Gladbach (Germany); Di Luna, L J; Moy, G [Teledyne Engineering Services, Waltham, MA (United States)

    1983-05-01

    This paper investigates plasticity, creep, and elastic follow-up effects on a full size hot primary piping system of the German fast breeder reactor prototype, the SNR-300. A large model (327 elements, 419 nodes) of straight pipe, special elbow and hanger elements of the general purpose finite element program, MARC-CDC, is used to predict piping behavior for a heat-up, sodium loading-unloading-reloading cycle and other significant operating conditions. Included in this work are many time-dependent solution increments for a 5,000 hour creep period. Creep strains and relaxed stress results, after 5,000 hours, for the complete model are used with uniaxial and biaxial models and results to extrapolate conclusions for a 100,000 hour operating life. (author)

  4. Inelastic analysis of SNR-300 piping

    International Nuclear Information System (INIS)

    Huebel, H.; Di Luna, L.J.; Moy, G.

    1983-01-01

    This paper investigates plasticity, creep, and elastic follow-up effects on a full size hot primary piping system of the German fast breeder reactor prototype, the SNR-300. A large model (327 elements, 419 nodes) of straight pipe, special elbow and hanger elements of the general purpose finite element program, MARC-CDC, is used to predict piping behavior for a heat-up, sodium loading-unloading-reloading cycle and other significant operating conditions. Included in this work are many time-dependent solution increments for a 5,000 hour creep period. Creep strains and relaxed stress results, after 5,000 hours, for the complete model are used with uniaxial and biaxial models and results to extrapolate conclusions for a 100,000 hour operating life. (author)

  5. Development of solutions to benchmark piping problems

    Energy Technology Data Exchange (ETDEWEB)

    Reich, M; Chang, T Y; Prachuktam, S; Hartzman, M

    1977-12-01

    Benchmark problems and their solutions are presented. The problems consist in calculating the static and dynamic response of selected piping structures subjected to a variety of loading conditions. The structures range from simple pipe geometries to a representative full scale primary nuclear piping system, which includes the various components and their supports. These structures are assumed to behave in a linear elastic fashion only, i.e., they experience small deformations and small displacements with no existing gaps, and remain elastic through their entire response. The solutions were obtained by using the program EPIPE, which is a modification of the widely available program SAP IV. A brief outline of the theoretical background of this program and its verification is also included.

  6. Physics of Earthquake Rupture Propagation

    Science.gov (United States)

    Xu, Shiqing; Fukuyama, Eiichi; Sagy, Amir; Doan, Mai-Linh

    2018-05-01

    A comprehensive understanding of earthquake rupture propagation requires the study of not only the sudden release of elastic strain energy during co-seismic slip, but also of other processes that operate at a variety of spatiotemporal scales. For example, the accumulation of the elastic strain energy usually takes decades to hundreds of years, and rupture propagation and termination modify the bulk properties of the surrounding medium that can influence the behavior of future earthquakes. To share recent findings in the multiscale investigation of earthquake rupture propagation, we held a session entitled "Physics of Earthquake Rupture Propagation" during the 2016 American Geophysical Union (AGU) Fall Meeting in San Francisco. The session included 46 poster and 32 oral presentations, reporting observations of natural earthquakes, numerical and experimental simulations of earthquake ruptures, and studies of earthquake fault friction. These presentations and discussions during and after the session suggested a need to document more formally the research findings, particularly new observations and views different from conventional ones, complexities in fault zone properties and loading conditions, the diversity of fault slip modes and their interactions, the evaluation of observational and model uncertainties, and comparison between empirical and physics-based models. Therefore, we organize this Special Issue (SI) of Tectonophysics under the same title as our AGU session, hoping to inspire future investigations. Eighteen articles (marked with "this issue") are included in this SI and grouped into the following six categories.

  7. Environmental hazards due to rupture of a liquefied propane pipeline

    International Nuclear Information System (INIS)

    Badr, O.A.; El-Sheikh, H.A.

    1996-01-01

    Accidental leakages of liquefied propane from high-pressure pipelines may occur despite the use of sophisticated safety equipment and following strict monitoring procedures. Environmental impact of steady and transient leakages were considered from toxicity and flammability viewpoints for two specific scenarios of full pipe ruptures. For each case, calculated mass flow rate, velocity, and temperature of leaking gas were utilized in an EPA-based dispersion model to predict the ground level concentration profiles in the downwind and crosswind directions. For the specific pipeline conditions considered here, the first scenario of a nonjet release (a cloud) produced steady toxic and flammable zones which were about 20 times bigger than those produced in the transient case. The second scenario of a free vertical jet resulted in the formation of a flammable vertical plume, while at ground level it did not produce flammable nor toxic zones. A parametric study of the first scenario confirmed the expected effects of both the gas release time and the atmospheric stability on the size of the dangerous zones. Within the typical range, the wind speed was found to have opposite effects for steady and transient releases. For a steady release, the dangerous zone was wider for slower winds and vice versa for a transient case. Moreover, the size of the dangerous zone was found to be an exponential function of the pipe diameter, while the effect of the initial pipe pressure was insignificant

  8. Development of prototype reactor maintenance. (2) Application to piping support of sodium-cooled reactor prototype

    International Nuclear Information System (INIS)

    Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji; Ito, Takaya; Yamaguchi, Akira

    2017-01-01

    A maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of piping supports could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports. (author)

  9. Steam generator tube rupture simulation using extended finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

    2016-08-15

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  10. Steam generator tube rupture simulation using extended finite element method

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

    2016-01-01

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  11. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

    2015-12-01

    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

  12. Heat-pipe Earth.

    Science.gov (United States)

    Moore, William B; Webb, A Alexander G

    2013-09-26

    The heat transport and lithospheric dynamics of early Earth are currently explained by plate tectonic and vertical tectonic models, but these do not offer a global synthesis consistent with the geologic record. Here we use numerical simulations and comparison with the geologic record to explore a heat-pipe model in which volcanism dominates surface heat transport. These simulations indicate that a cold and thick lithosphere developed as a result of frequent volcanic eruptions that advected surface materials downwards. Declining heat sources over time led to an abrupt transition to plate tectonics. Consistent with model predictions, the geologic record shows rapid volcanic resurfacing, contractional deformation, a low geothermal gradient across the bulk of the lithosphere and a rapid decrease in heat-pipe volcanism after initiation of plate tectonics. The heat-pipe Earth model therefore offers a coherent geodynamic framework in which to explore the evolution of our planet before the onset of plate tectonics.

  13. Spontaneous rupture of vaginal enterocele

    DEFF Research Database (Denmark)

    Svendsen, J H; Galatius, H; Hansen, P K

    1985-01-01

    Spontaneous rupture of an enterocele is a rare complication. Only 24 cases including the present case have been reported in the literature. The patients were elderly and had had at least one vaginal operation. The patients were remarkably unaffected symptomatically on admission.......Spontaneous rupture of an enterocele is a rare complication. Only 24 cases including the present case have been reported in the literature. The patients were elderly and had had at least one vaginal operation. The patients were remarkably unaffected symptomatically on admission....

  14. Rupture luminescence from natural fibers

    Science.gov (United States)

    Li, W.; Haneman, D.

    1999-12-01

    Fibers of cotton and wool, and samples of paper, have been ruptured in tension in vacuum and in air, and give detectable luminescence in the visible range. All have a common emission peak at around 2.0 eV, which is ascribed to the deexcitation of states excited by the rupture of organic chain molecule bonds. Rubber bands give stronger emission in air, but no emission in vacuum, suggesting the material breaks only at weak interchain bonds. Mohair, cat, and horse hair also give emission in air. The phenomena reveal effects that would occur widely in nature.

  15. 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)

  16. Heat pipes and use of heat pipes in furnace exhaust

    Science.gov (United States)

    Polcyn, Adam D.

    2010-12-28

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

  17. Analysis of 30 breast implant rupture cases.

    Science.gov (United States)

    Tark, Kwan Chul; Jeong, Hii Sun; Roh, Tae Suk; Choi, Jong Woo

    2005-01-01

    Breast implants used for augmentation mammoplasty or breast reconstruction could rupture from various causes such as trauma or spontaneous failure. The objectives of this study were to investigate the relationships between the causes of implant rupture and the degree of capsular contracture, and then to evaluate the relative efficacies of specific signs on magnetic resonance imaging (MRI) known to be beneficial for diagnosing the rupture. A retrospective review identified patients with prosthetic implant rupture or impending rupture treated by the senior author. The 30 cases of implant rupture available for review were classified into two groups: intracapsular and extracapsular ruptures. The 30 cases of breast implant ruptures were analyzed with respect to the clinical symptoms and signs, the causes of rupture, the degree of capsular contracture, and therapeutic plans. Among the 30 cases, 14 patients who had undergone MRI during the diagnostic period were analyzed with respect to the relationships between MRI readings and operative findings. Spontaneous rupture of membranes was most common (80%), followed by failure because of trauma (7%) and valve or implant base (4%). The symptoms during implant rupture were contour deformity, palpated mass-like lesions, pain, and focal inflammation. According to the analysis of specific MRI signs, the sensitivity and specificity of the linguine sign were 87% and 100%, respectively, for intracapsular rupture. For extracapsular rupture, the sensitivity and specificity of the linguine sign were, respectively, 67% and 75%. The sensitivity and specificity of the rat-tail sign and tear drop sign were 14% and 50%, respectively. Breast implant rupture was correlated with the degree of capsular contracture in our study. Among the various specific MRI signs used in diagnosing the rupture, the linguine sign was reliable and had a high sensitivity and specificity, especially in cases of intracapsular rupture. On the other hand, the rat

  18. PE 100 pipe systems

    CERN Document Server

    Brömstrup, Heiner

    2012-01-01

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

  19. Rupture of Achilles Tendon : Usefulness of Ultrasonography

    International Nuclear Information System (INIS)

    Kim, Nam Hyeon; Ki, Won Woo; Yoon, Kwon Ha; Kim, Song Mun; Shin, Myeong Jin; Kwon, Soon Tae

    1996-01-01

    To differentiate a complete rupture of Achilles tendon from an incomplete one which is important because its treatment is quite different. And it is necessary to know the exact site of the rupture preoperatively. Fifteen cases of fourteen patients which were diagnosed as Achilles tendon rupture by ultrasonography and surgery were reviewed. We compared sonographic rupture site with surgical findings. Ultrasonographic criteria for differentiation of complete and incomplete rupture was defined as follows : the discreteness, which means the proximal intervening hypoechogenicity to the interface echogenicity of distal margin of ruptured tendon : the slant sign, which represents the interface of ruptured distal margin which was seen over the 3/4 of the thickness of the tendon without intervening low echogeneicity : the invagination sign, which means the echogenic invagination from Kager triangle into posterior aspect of Achilles tendon over the half thickness of the tendon. The sites of complete tendon rupture were exactly corresponded to surgical finding in four cases of ten complete ruptures. And the discrepancy between sonographic and surgical findings in the site of complete rupture was 1.2 ± 0.4 cm in six cases. Three of ten complete ruptures showed the discreteness sign, all of ten showed the slant sign and two of ten showed the invagination sign. It is helpful to differentiate a complete from incomplete rupture of the Achilles tendon and to localize the site of the complete rupture with the ultrasonographic evaluation

  20. 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)

  1. Pipe-to-pipe impact tests

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-11-01

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

  2. Performance of buried pipe installation.

    Science.gov (United States)

    2010-05-01

    The purpose of this study is to determine the effects of geometric and mechanical parameters : characterizing the soil structure interaction developed in a buried pipe installation located under : roads/highways. The drainage pipes or culverts instal...

  3. U.S. Nuclear Regulatory Commission Extremely Low Probability of Rupture pilot study : xLPR framework model user's guide.

    Energy Technology Data Exchange (ETDEWEB)

    Kalinich, Donald A.; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01

    For the U.S. Nuclear Regulatory Commission (NRC) Extremely Low Probability of Rupture (xLPR) pilot study, Sandia National Laboratories (SNL) was tasked to develop and evaluate a probabilistic framework using a commercial software package for Version 1.0 of the xLPR Code. Version 1.0 of the xLPR code is focused assessing the probability of rupture due to primary water stress corrosion cracking in dissimilar metal welds in pressurizer surge nozzles. Future versions of this framework will expand the capabilities to other cracking mechanisms, and other piping systems for both pressurized water reactors and boiling water reactors. The goal of the pilot study project is to plan the xLPR framework transition from Version 1.0 to Version 2.0; hence the initial Version 1.0 framework and code development will be used to define the requirements for Version 2.0. The software documented in this report has been developed and tested solely for this purpose. This framework and demonstration problem will be used to evaluate the commercial software's capabilities and applicability for use in creating the final version of the xLPR framework. This report details the design, system requirements, and the steps necessary to use the commercial-code based xLPR framework developed by SNL.

  4. PREMATURE RUPTURE OF THE MEMBRANES*

    African Journals Online (AJOL)

    In patients presenting with premature rupture of the membranes there are two factors which influence the foetal morbidity and mortality. These factors are prema- turity and intra-uterine infection. The purpose of this analysis was to elucidate which factor carried the greater risk to the foetus. Recently there has been a spate of.

  5. Ruptured Cervical Anellrysm with Neurofibromatosis

    African Journals Online (AJOL)

    rupture of an aneurysm of a lateral branch of the right thyrocervical trunk in a patient suffering from diffuse neurofibromatosis. The operative findings are reported. s. Air. Med. ... system demonstrated decreased power in all muscles of the right shoulder girdle and arm, without sensory change. The cranial nerves were normal.

  6. Spontaneous Splenic Rupture in Melanoma

    Directory of Open Access Journals (Sweden)

    Hadi Mirfazaelian

    2014-01-01

    Full Text Available Spontaneous rupture of spleen due to malignant melanoma is a rare situation, with only a few case reports in the literature. This study reports a previously healthy, 30-year-old man who came with chief complaint of acute abdominal pain to emergency room. On physical examination, abdominal tenderness and guarding were detected to be coincident with hypotension. Ultrasonography revealed mild splenomegaly with moderate free fluid in abdominopelvic cavity. Considering acute abdominal pain and hemodynamic instability, he underwent splenectomy with splenic rupture as the source of bleeding. Histologic examination showed diffuse infiltration by tumor. Immunohistochemical study (positive for S100, HMB45, and vimentin and negative for CK, CD10, CK20, CK7, CD30, LCA, EMA, and chromogranin confirmed metastatic malignant melanoma. On further questioning, there was a past history of a nasal dark skin lesion which was removed two years ago with no pathologic examination. Spontaneous (nontraumatic rupture of spleen is an uncommon situation and it happens very rarely due to neoplastic metastasis. Metastasis of malignant melanoma is one of the rare causes of the spontaneous rupture of spleen.

  7. Endometriosis-related spontaneous diaphragmatic rupture.

    Science.gov (United States)

    Triponez, Frédéric; Alifano, Marco; Bobbio, Antonio; Regnard, Jean-François

    2010-10-01

    Non-traumatic, spontaneous diaphragmatic rupture is a rare event whose pathophysiology is not known. We report the case of endometriosis-related spontaneous rupture of the right diaphragm with intrathoracic herniation of the liver, gallbladder and colon. We hypothesize that the invasiveness of endometriotic tissue caused diaphragm fragility, which finally lead to its complete rupture without traumatic event. The treatment consisted of a classical management of diaphragmatic rupture, with excision of the endometriotic nodule followed by medical ovarian suppression for six months.

  8. Optimization of Pipe Networks

    DEFF Research Database (Denmark)

    Hansen, C. T.; Madsen, Kaj; Nielsen, Hans Bruun

    1991-01-01

    algorithm using successive linear programming is presented. The performance of the algorithm is illustrated by optimizing a network with 201 pipes and 172 nodes. It is concluded that the new algorithm seems to be very efficient and stable, and that it always finds a solution with a cost near the best...

  9. ACL Rupture in Collegiate Wrestler

    Directory of Open Access Journals (Sweden)

    Lindsay A. Palmer

    2016-05-01

    Full Text Available Objective: To educate others on unique Anterior Cruciate Ligament tears and percentage of usage of the ACL in normal daily function. Background: Patient is an eighteen year old male participating in wrestling and football at the time of the injury. Patient now only participates in wrestling. No previous knee or chronic injuries were reported prior to this injury. Patient was playing football during the time of injury. The patient stated that he planted his foot down and was tackled at the same time when the injury occurred. The patient felt his knee twist and buckle. Patient complained of clicking inside the knee and had minimal swelling. He also complained of it being difficult to bear weight at the time. The patient did not seek further treatment until two months after the injury occurred when he received an MRI. His MRI showed a positive finding for an Anterior Cruciate Ligament rupture. His previous Athletic Trainer could not find a positive diagnosis for the patient prior to the MRI. Differential Diagnosis: Possible meniscal or ACL injury. Treatment: Doctors officially diagnosed the injury as a complete rupture of the ACL. The patient did not receive surgery immediately. Doctors have stated that he only uses about 50% of his ACL on a daily basis compared to a normal person who uses about 95% of their ACL daily. Because of this, the patient played on his rupture for seven months before receiving surgery. He played a whole season of high school football and a whole season of wrestling his senior year with the ACL ruptured. The patient only used a brace for better comfort during the seven months. The patient then received reconstructive surgery to repair the rupture. A hamstring tendon graft was used to repair the ruptured ACL. Because a tendon was taken from the hamstring, patient experienced a tight ACL and hamstring of the left leg post-surgery. The patient participated in Physical Therapy for five months to strengthen and stretch the new

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

    International Nuclear Information System (INIS)

    Jiang Zhenbiao; Li Guanying; Liu Zhicheng

    2013-01-01

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

  11. A Retrospective Analysis of Ruptured Breast Implants

    Directory of Open Access Journals (Sweden)

    Woo Yeol Baek

    2014-11-01

    Full Text Available BackgroundRupture is an important complication of breast implants. Before cohesive gel silicone implants, rupture rates of both saline and silicone breast implants were over 10%. Through an analysis of ruptured implants, we can determine the various factors related to ruptured implants.MethodsWe performed a retrospective review of 72 implants that were removed for implant rupture between 2005 and 2014 at a single institution. The following data were collected: type of implants (saline or silicone, duration of implantation, type of implant shell, degree of capsular contracture, associated symptoms, cause of rupture, diagnostic tools, and management.ResultsForty-five Saline implants and 27 silicone implants were used. Rupture was diagnosed at a mean of 5.6 and 12 years after insertion of saline and silicone implants, respectively. There was no association between shell type and risk of rupture. Spontaneous was the most common reason for the rupture. Rupture management was implant change (39 case, microfat graft (2 case, removal only (14 case, and follow-up loss (17 case.ConclusionsSaline implants have a shorter average duration of rupture, but diagnosis is easier and safer, leading to fewer complications. Previous-generation silicone implants required frequent follow-up observation, and it is recommended that they be changed to a cohesive gel implant before hidden rupture occurs.

  12. Venting of gas deflagrations through relief pipes

    OpenAIRE

    Ferrara, Gabriele

    2006-01-01

    Vent devices for gas and dust explosions are often ducted to safety locations by means of relief pipes for the discharge of hot combustion products or blast waves (NFPA 68, 2002). The presence of the duct is likely to increase the severity of the explosion with respect to simply vented vessels posing a problem for the proper design of this venting configuration. The phenomenology of the vented explosion is complicated as the interaction of combustion in the duct with primary combustion in...

  13. 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.

  14. Identification of significant problems related to light water reactor piping systems

    International Nuclear Information System (INIS)

    1980-07-01

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

  15. A Markov chain model for CANDU feeder pipe degradation

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. Heat pipe applications workshop report

    International Nuclear Information System (INIS)

    Ranken, W.A.

    1978-04-01

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

  17. Piping reliability analysis: Some views on the roles of data-driven models vs. probabilistic fracture mechanics (PFM)

    International Nuclear Information System (INIS)

    Lydell, B.

    1997-01-01

    The objective of the presentation is to address the question in five different perspectives: Historical; Methodological; Quality PSA and the specifications for pipe rupture frequency estimation -verification and validation; User-perspectives on frequency estimation; data analysis perspectives on the choice of estimation technique

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  19. Diffusion in flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brogaard Kristensen, S.

    2000-06-01

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

  20. Diffusion in flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brogaard Kristensen, S

    2000-06-01

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

  1. Piping failures in United States nuclear power plants 1961-1995

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. Acute longitudinal ligament rupture following acute spinal trauma

    Directory of Open Access Journals (Sweden)

    Donald Hansom

    2014-06-01

    Full Text Available The authors present a rare case of anterior longitudinal ligament (ALL rupture in a 47- year-old gentleman following a bicycle accident. The ALL is a continuous band of a variable thickness that acts as a primary spinal stabiliser. Stress, strain or rupture of the ALL usually occurs as a result of hyperextension, with the primary perpetrator being whiplash injuries. Such injuries have been shown to result in cervical spine instability during extension, axial rotation, and lateral bending modes. Spine radiographs of such patients may be routinely assessed as normal, therefore this specific type of injury does not lend itself to identification by traditional imaging methods. This account demonstrates the importance of having a high index of suspicion of a ligamentous neck injury in the setting of normal plain radiographs but abnormal clinical examination.

  3. Achilles tendon rupture in badminton.

    Science.gov (United States)

    Kaalund, S; Lass, P; Høgsaa, B; Nøhr, M

    1989-01-01

    The typical badminton player with an Achilles tendon rupture is 36 years old and, despite limbering up, is injured at the rear line in a sudden forward movement. He resumes work within three months and has a slight lack of dorsiflexion in the ankle as the main complication. Most patients resume badminton within one year, but some finish their sports career, mainly due to fear of a new injury. The investigation discusses predisposing factors and prophylactic measures. PMID:2605439

  4. Simultaneous uterine and urinary bladder rupture in an otherwise successful vaginal birth after cesarean delivery.

    Science.gov (United States)

    Ho, Szu-Ying; Chang, Shuenn-Dhy; Liang, Ching-Chung

    2010-12-01

    Uterine rupture is the primary concern when a patient chooses a trial of labor after a cesarean section. Bladder rupture accompanied by uterine rupture should be taken into consideration if gross hematuria occurs. We report the case of a patient with uterine rupture during a trial of labor after cesarean delivery. She had a normal course of labor and no classic signs of uterine rupture. However, gross hematuria was noted after repair of the episiotomy. The patient began to complain of progressive abdominal pain, gross hematuria and oliguria. Cystoscopy revealed a direct communication between the bladder and the uterus. When opening the bladder peritoneum, rupture sites over the anterior uterus and posterior wall of the bladder were noted. Following primary repair of both wounds, a Foley catheter was left in place for 12 days. The patient had achieved a full recovery by the 2-year follow-up examination. Bladder injury and uterine rupture can occur at any time during labor. Gross hematuria immediately after delivery is the most common presentation. Cystoscopy is a good tool to identify the severity of bladder injury. Copyright © 2010 Elsevier. Published by Elsevier B.V. All rights reserved.

  5. Simultaneous Uterine and Urinary Bladder Rupture in an Otherwise Successful Vaginal Birth After Cesarean Delivery

    Directory of Open Access Journals (Sweden)

    Szu-Ying Ho

    2010-12-01

    Full Text Available Uterine rupture is the primary concern when a patient chooses a trial of labor after a cesarean section. Bladder rupture accompanied by uterine rupture should be taken into consideration if gross hematuria occurs. We report the case of a patient with uterine rupture during a trial of labor after cesarean delivery. She had a normal course of labor and no classic signs of uterine rupture. However, gross hematuria was noted after repair of the episiotomy. The patient began to complain of progressive abdominal pain, gross hematuria and oliguria. Cystoscopy revealed a direct communication between the bladder and the uterus. When opening the bladder peritoneum, rupture sites over the anterior uterus and posterior wall of the bladder were noted. Following primary repair of both wounds, a Foley catheter was left in place for 12 days. The patient had achieved a full recovery by the 2-year follow-up examination. Bladder injury and uterine rupture can occur at any time during labor. Gross hematuria immediately after delivery is the most common presentation. Cystoscopy is a good tool to identify the severity of bladder injury.

  6. Waste pipe calculus

    International Nuclear Information System (INIS)

    Kaufman, A.M.

    1978-01-01

    A rapid method is presented for calculating transport in a network of one-dimensional flow paths or ''pipes''. The method defines a Green's function for each flow path and prescribes a method of combining these Green's functions to produce an overall Green's function for the flow path network. A unique feature of the method is the use of the Laplace transform of these Green's functions to carry out most of the calculations

  7. 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

  8. High-grade Angiosarcoma Associated with Ruptured Breast Implants

    Directory of Open Access Journals (Sweden)

    Nicolas R. Smoll, MBBS

    2013-04-01

    Full Text Available Summary: Since the serendipitous discovery that implanted polymers cause sarcomas in rats, much research has been conducted to prove or disprove a link between silicone breast implants and/or polymer-based materials and breast cancer. In light of an initial report that 35% of rats implanted with a variety of polymers developed fibrosarcomas, we report a case of primary angiosarcoma found in a patient presenting with bilateral rupture of gel-filled breast implants.

  9. 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

  10. Heat pipes and heat pipe exchangers for heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-01-01

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

  11. 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.

  12. Complex rupture during the 12 January 2010 Haiti earthquake

    Science.gov (United States)

    Hayes, G.P.; Briggs, R.W.; Sladen, A.; Fielding, E.J.; Prentice, C.; Hudnut, K.; Mann, P.; Taylor, F.W.; Crone, A.J.; Gold, R.; Ito, T.; Simons, M.

    2010-01-01

    Initially, the devastating Mw 7.0, 12 January 2010 Haiti earthquake seemed to involve straightforward accommodation of oblique relative motion between the Caribbean and North American plates along the Enriquillog-Plantain Garden fault zone. Here, we combine seismological observations, geologic field data and space geodetic measurements to show that, instead, the rupture process may have involved slip on multiple faults. Primary surface deformation was driven by rupture on blind thrust faults with only minor, deep, lateral slip along or near the main Enriquillog-Plantain Garden fault zone; thus the event only partially relieved centuries of accumulated left-lateral strain on a small part of the plate-boundary system. Together with the predominance of shallow off-fault thrusting, the lack of surface deformation implies that remaining shallow shear strain will be released in future surface-rupturing earthquakes on the Enriquillog-Plantain Garden fault zone, as occurred in inferred Holocene and probable historic events. We suggest that the geological signature of this earthquakeg-broad warping and coastal deformation rather than surface rupture along the main fault zoneg-will not be easily recognized by standard palaeoseismic studies. We conclude that similarly complex earthquakes in tectonic environments that accommodate both translation and convergenceg-such as the San Andreas fault through the Transverse Ranges of Californiag-may be missing from the prehistoric earthquake record. ?? 2010 Macmillan Publishers Limited. All rights reserved.

  13. Total transverse rupture of the duodenum after blunt abdominal trauma.

    Science.gov (United States)

    Pirozzi, Cesare; Di Marco, Carluccio; Loponte, Margherita; Savino, Grazia

    2014-05-11

    Complete transverse rupture of the duodenum as an isolated lesion in blunt trauma can be considered as exceptional. The aim of this report is to discuss diagnostic procedures and surgical options in such an infrequent presentation. We report on a 37 year old man who had a total transverse rupture of the duodenum after blunt abdominal trauma. Diagnosis was suspected after contrast enhanced CT scan and confirmed at laparotomy; duodenal rupture was repaired by an end to end duodenal-duodenal anastomosis, after Kocher maneuver. The patient had fast and complete recovery. A high index of suspicion is necessary for timely diagnosis. Multi detector contrast enhanced CT scan is the gold standard for that aim. Surgical management must be tailored on an individual basis, since many techniques are available for both reconstruction and duodenum decompression. Kocher maneuver is essential for complete inspection of the pancreatic duodenal block and for appropriate reconstruction. Management of isolated duodenal rupture can be difficult. Contrast enhanced TC scans is essential for timely diagnosis. Primary repair can be achieved by an end to end duodenum anastomosis after Kocher maneuver, although alternative techniques are available for tailored solutions. Complex duodenum decompression techniques are not mandatory.

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

    International Nuclear Information System (INIS)

    Kim, Taesoon; Lee, Dohwan

    2016-01-01

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

  15. Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. Fatigue check of nuclear safety class 1 reactor coolant pipe

    International Nuclear Information System (INIS)

    Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

    2015-01-01

    Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

  17. Fatigue of LMFBR piping due to flow stratification

    Energy Technology Data Exchange (ETDEWEB)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface.

  18. Fatigue of LMFBR piping due to flow stratification

    International Nuclear Information System (INIS)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface

  19. A case for historic joint rupture of the San Andreas and San Jacinto faults

    OpenAIRE

    Lozos, Julian C.

    2016-01-01

    The San Andreas fault is considered to be the primary plate boundary fault in southern California and the most likely fault to produce a major earthquake. I use dynamic rupture modeling to show that the San Jacinto fault is capable of rupturing along with the San Andreas in a single earthquake, and interpret these results along with existing paleoseismic data and historic damage reports to suggest that this has likely occurred in the historic past. In particular, I find that paleoseismic data...

  20. Pipe support program at Pickering

    International Nuclear Information System (INIS)

    Sahazizian, L.A.; Jazic, Z.

    1997-01-01

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

  1. Acoustic analysis of a piping system

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. 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

  3. 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

  4. Water hammer in elastic pipes

    International Nuclear Information System (INIS)

    Gale, J.; Tiselj, I.

    2002-01-01

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

  5. Plastics pipe couplings

    International Nuclear Information System (INIS)

    Glover, J.B.

    1980-07-01

    A method is described of making a pipe coupling of the type comprising a plastics socket and a resilient annular sealing member secured in the mouth thereof, in which the material of at least one component of the coupling is subjected to irradiation with high energy radiation whereby the material is caused to undergo cross-linking. As examples, the coupling may comprise a polyethylene or plasticised PVC socket the material of which is subjected to irradiation, and the sealing member may be moulded from a thermoplastic elastomer which is subjected to irradiation. (U.K.)

  6. Rehabilitation of the gas pipeline that had a rupture in service caused by SCC (Stress Corrosion C raking); Rehabilitacion al servicio de un gasoducto que ha sufrido una ruptura en servicio por SCC

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fernando; Carzoglio, Eduardo; Hryciuk, Pedro [TGN - Transportadora de Gas del Norte S.A. (Argentina). Depto. de Integridad

    2003-07-01

    TGN had a rupture in service on Gasoducto Troncal Norte. After initial evaluation of the causes of the rupture it was concluded that it had been caused by Stress Corrosion Cracking (SCC). Subsequent investigation in the area of the rupture revealed that colonies of cracks, typical of SCC were found in pipes located near the rupture. In order to put back in service the pipeline in a safety condition, SCC mitigation activities were performed. A decision was made to conduct a hydro test along approximately 30 kilometers of pipe. The stages of the works, the problems faced and the solutions found are dealt with, as well as the conclusions reached upon completion of the works which allowed a better understanding of SCC phenomenon. The methodology for the identification of those areas susceptible to SCC is also described. (author)

  7. Neck curve polynomials in neck rupture model

    International Nuclear Information System (INIS)

    Kurniadi, Rizal; Perkasa, Yudha S.; Waris, Abdul

    2012-01-01

    The Neck Rupture Model is a model that explains the scission process which has smallest radius in liquid drop at certain position. Old fashion of rupture position is determined randomly so that has been called as Random Neck Rupture Model (RNRM). The neck curve polynomials have been employed in the Neck Rupture Model for calculation the fission yield of neutron induced fission reaction of 280 X 90 with changing of order of polynomials as well as temperature. The neck curve polynomials approximation shows the important effects in shaping of fission yield curve.

  8. Acute Pectoralis Major Rupture Captured on Video

    Directory of Open Access Journals (Sweden)

    Alejandro Ordas Bayon

    2016-01-01

    Full Text Available Pectoralis major (PM ruptures are uncommon injuries, although they are becoming more frequent. We report a case of a PM rupture in a young male who presented with axillar pain and absence of the anterior axillary fold after he perceived a snap while lifting 200 kg in the bench press. Diagnosis of PM rupture was suspected clinically and confirmed with imaging studies. The patient was treated surgically, reinserting the tendon to the humerus with suture anchors. One-year follow-up showed excellent results. The patient was recording his training on video, so we can observe in detail the most common mechanism of injury of PM rupture.

  9. Traumatic rupture of an intracranial dermoid cyst

    Directory of Open Access Journals (Sweden)

    Raksha Ramlakhan, BMedSc, MBBCh

    2015-01-01

    Full Text Available Intracranial dermoid cysts are congenital tumors of ectodermal origin. Rupture of these cysts can occur spontaneously, but rupture in association with trauma is reported infrequently. The diagnosis of rupture is made by the presence of lipid (cholesterol droplets in the subarachnoid spaces and ventricles. Nonenhanced CT of the head demonstrates multiple foci of low attenuation that correspond with hyperintense signal on T1-weighted MRI. We present a case of an adult patient with rupture of an intracranial dermoid cyst, precipitated by minor trauma.

  10. CT diagnosis of ruptured abdominal aortic aneurysm

    International Nuclear Information System (INIS)

    Sacknoff, R.; Novelline, R.A.; Wittenberg, J.; Waltman, A.C.; De Luca, S.A.; Rhea, J.T.; Lawrason, J.N.

    1986-01-01

    Ruptured abdominal aortic aneurysm (AAA) is a life-threatening condition requiring immediate diagnosis and surgery. In a series of 23 consecutive patients scanned by CT for suspected ruptured AAA, CT proved 100% accurate. In seven patients with surgically or pathologically proved ruptured AAA, CT demonstrated a similar distribution of hemorrhage into the perirenal space and to a lesser degree into the anterior and posterior pararenal spaces. The 16 true-negative examinations included ten in patients with unruptured AAA and six in patients with other diseases. The authors conclude that patients in stable condition with suspected ruptured AAA should be examined by CT

  11. 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

  12. Heat pipe and method of production of a heat pipe

    International Nuclear Information System (INIS)

    Kemp, R.S.

    1975-01-01

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

  13. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  14. LMFBR flexible pipe joint development program. Annual technical progress report, government fiscal year 1977

    International Nuclear Information System (INIS)

    1978-01-01

    Currently, the ASME Boiler and Pressure Vessel Code does not allow the use of flexible pipe joints (bellows) in Section III, Class 1 reactor primary piping systems. Studies have shown that the primary piping loops of LMFBR's could be simplified by using these joints. This simplification translates directly into shorter primary piping runs and reduced costs for the primary piping system. Further cost savings result through reduced vault sizes and reduced containment building diameter. In addition, the use of flexible joints localizes the motions from thermally-induced piping growth into components which are specifically designed to accommodate this motion. This reduces the stress levels in the piping system and its components. It is thus economically and structurally important that flexible piping joints be available to the LMFBR designer. The overall objective of the Flexible Joint Program is to provide this availability. This will be accomplished through the development of ASME rules which allow the appropriate use of such joints in Section III, Class 1 piping systems and through the development and demonstration of construction methods which satisfy these rules. The rule development includes analytic and testing methodology formulations which will be supported by subscale bellows testing. The construction development and demonstration encompass the design, fabrication, and in-sodium testing of prototypical LMFBR plant-size flexible pipe joints which meet all ASME rule requirements. The satisfactory completion of these developmental goals will result in an approved flexible pipe joint design for the LMFBR. Progress is summarized in the following efforts undertaken during 1977 to accomplish these goals: (1) code case support, (2) engineering and design, (3) material development, (4) testing, and (5) manufacturing development

  15. Mechanical Behaviour of Lined Pipe

    NARCIS (Netherlands)

    Hilberink, A.

    2011-01-01

    Installing lined pipe by means of the reeling installation method seems to be an attractive combination, because it provides the opportunity of eliminating the demanding welds from the critical time offshore and instead preparing them onshore. However, reeling of lined pipe is not yet proven

  16. Pulsed TIG welding of pipes

    International Nuclear Information System (INIS)

    Killing, U.

    1989-01-01

    The present study investigates into the effects of impulse welding parameters on weld geometry in the joint welding of thin-walled sheets and pipes (d=2.5 mm), and it uses random samples of thick-walled sheets and pipes (d=10 mm), in fixed positions. (orig./MM) [de

  17. Effect of initial fluid-system pressures on the behavior of a rupture-disc pressure-relief device

    International Nuclear Information System (INIS)

    Hsieh, B.J.; Shin, Y.W.; Kot, C.A.

    1983-01-01

    Rupture disc assemblies are used in piping network systems as a pressure-relief device to protect the system from being exposed to excess pressures. Among the various disc assemblies, the reverse-buckling type is chosen for application in the Clinch River Breeder Reactor. This rupture-disc assembly consists of a portion of a thin spherical shell with its convex side subjected to the fluid system. The reverse-buckling type rupture disc assemblies have been used successfully in environments where the fluid is gas, i.e. highly compressible, and their performances have been judged as adequate in the liquid environment. To analyze the piping system, an analysis method is needed taking into consideration of the fluid/disc interaction, the nonlinear dynamic buckling phenomenon of the disc, and the possible cavitation of the fluid. A computer code SWAAM-I had been written at the Components Technology Division, Argonne National Laboratory. Among its many functions, one is to compute the response of 1-dimensional pressure pulse propagation including the effects of many different types of boundary conditions and possible pipe plasticity

  18. Representative stresses for creep deformation and failure of pressurised tubes and pipes

    International Nuclear Information System (INIS)

    Cane, B.J.; Browne, R.J.

    1982-01-01

    Results of a series of tube and uniaxial tests on two casts of 1/2CrMoV pipe steel are examined to determine the representative stress which must be applied to uniaxial data in order to predict the strain rates and lives of pressurised tubes and pipes. The stress criterion for deformation is shown to correlate with the analytically derived reference stress (σsub(R)) at low pressure while at high pressures a modified reference stress (> σsub(R)) must be used. The rupture life exhibits a similar correlation such that the representative stress for rupture is given by σsub(R) at low stresses yet, at high stresses, it is greater than σsub(R) and attains a value comparable with the mean diameter hoop stress. The latter thus describes the rupture life at high pressures but significantly underestimates the life at low pressures approaching those in service. Consideration is given to the multiaxial stress rupture criterion and the effect of geometry in constant load tests. (author)

  19. Functional capability of piping systems

    International Nuclear Information System (INIS)

    Terao, D.; Rodabaugh, E.C.

    1992-11-01

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

  20. Promethus Hot Leg Piping Concept

    International Nuclear Information System (INIS)

    AM Girbik; PA Dilorenzo

    2006-01-01

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

  1. FFTF report: FFTF piping installation and welding techniques

    International Nuclear Information System (INIS)

    Gilles, J.

    1975-01-01

    The main sodium piping with a diameter of 16'' or 28 '' is being installed at the FFTF construction site starting in December 1974. The supplier and authority demarcations are: Combustion Engineering supplies the reactor vessel, guard vessel and adjoining pipes and uses the machine welding equipment ''Dimetrics''; for the piping system of the primary and secondary loops the pipes manufactured by Rollmet at HUICO, Pasco, were delivered and prefabricated there, as far as compatible with the installation. ''Astroarc'' welding machines are used by Bechtel for the piping prefabrication in the weld laboratory as well as on site at the construction site. Technical welding problems occurring during the course of the installation at the construction site and several during this time are described. At present 6 weld seams in the reactor and 14 weld seams in the secondary loop are accepted. The requirement exists to carry out as many welds as possible automatically, in order to produce sodium pipe welds of high technical quality and which are reproducible. The welding equipment is described

  2. 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.)

  3. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    International Nuclear Information System (INIS)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik; Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae

    2015-01-01

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  4. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik [School of Materials Science and Engineering, Andong National University, Andong (Korea, Republic of); Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae [Power Engineering Research Institute, KEPCO Engineering and Construction Company, Seongnam (Korea, Republic of)

    2015-02-15

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  5. Specialist meeting on leak before break in reactor piping and vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  6. 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

  7. 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

  8. Investigation of erosion behavior in different pipe-fitting using Eulerian-Lagrangian approach

    Science.gov (United States)

    Kulkarni, Harshwardhan; Khadamkar, Hrushikesh; Mathpati, Channamallikarjun

    2017-11-01

    Erosion is a wear mechanism of piping system in which wall thinning occurs because of turbulent flow along with along with impact of solid particle on the pipe wall, because of this pipe ruptures causes costly repair of plant and personal injuries. In this study two way coupled Eulerian-Lagrangian approach is used to solve the liquid solid (water-ferrous suspension) flow in the different pipe fitting namely elbow, t-junction, reducer, orifice and 50% open gate valve. Simulations carried out using incomressible transient solver in OpenFOAM for different Reynolds's number (10k, 25k, 50k) and using WenYu drag model to find out possible higher erosion region in pipe fitting. Used transient solver is a hybrid in nature which is combination of Lagrangian library and pimpleFoam. Result obtained from simulation shows that exit region of elbow specially downstream of straight, extradose of the bend section more affected by erosion. Centrifugal force on solid particle at bend affect the erosion behavior. In case of t-junction erosion occurs below the locus of the projection of branch pipe on the wall. For the case of reducer, orifice and a gate valve reduction area as well as downstream is getting more affected by erosion because of increase in velocities.

  9. 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.

  10. Nonlinear fluid/structure interaction relating a rupture-disc pressure-relief device

    International Nuclear Information System (INIS)

    Hsieh, B.J.; Kot, C.A.; Shin, Y.W.; Youngdahl, C.K.

    1983-01-01

    Rupture disc assemblies are used in piping network systems as a pressure-relief device. The reverse-buckling type is chosen for application in a liquid metal fast breeder reactor. This assembly is used successfully in systems in which the fluid is highly compressible, such as air; the opening up of the disc by the knife setup is complete. However, this is not true for a liquid system; it had been observed experimentally that the disc may open up only partially or not at all. Therefore, to realistically understand and represent a rupture disc assembly in a liquid environment, the fluid-structure interactions between the liquid medium and the disc assembly must be considered. The methods for analyzing the fluid and the disc and the mechanism interconnecting them are presented. The fluid is allowed to cavitate through a column-cavitation model and the disc is allowed to become plastically deformed through the classic Von Mises' yield criteria, when necessary

  11. Magnetic non-destructive evaluation of ruptures of tensile armor in oil risers

    International Nuclear Information System (INIS)

    Pérez-Benitez, J A; Padovese, L R

    2012-01-01

    Risers are flexible multilayered pipes formed by an inner flexible metal structure surrounded by polymer layers and spiral wound steel ligaments, also known as armor wires. Since these risers are used to link subsea pipelines to floating oil and gas production installations, and their failure could produce catastrophic consequences, some methods have been proposed to monitor the armor integrity. However, until now there is no practical method that allows the automatic non-destructive detection of individual armor wire rupture. In this work we show a method using magnetic Barkhausen noise that has shown high efficiency in the detection of armor wire rupture. The results are examined under the cyclic and static load conditions of the riser. This work also analyzes the theory behind the singular dependence of the magnetic Barkhausen noise on the applied tension in riser armor wires. (paper)

  12. Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Larson, T.K.

    1978-01-01

    Examination of the U-tubes in the steam generators of some large commercial pressurized water reactors (PWR) has revealed the existence of leakage and in some cases structural weakening of the tubes. This structural weakening enhances the possibility of tubes rupturing during a hypothesized loss-of-coolant accident (LOCA). Considerable interest has been shown in the analysis of tube ruptures concurrent with a hypothesized LOCA since the presence of tube ruptures has the potential to influence the system thermal-hydraulic response and could foreseeably result in a more severe core thermal behavior than might otherwise occur. To experimentally investigate the influence of steam generator tube ruptures on the thermal-hydraulic response of PWR type system, a series of experiments was conducted in the Semiscale Mod-1 system by EG and G Idaho, Inc., for the U.S. Nuclear Regulatory Commission and the Department of Energy. The primary objective of the experiments was to obtain data which could be used to evaluate the influence of the simulated tube ruptures on the system and core thermal-hydraulic response for a range of tube ruptures that was expected to provide the potential for high cladding temperatures in the Semiscale facility. The experiments were conducted assuming a variety in the number of tubes ruptured during large break loss-of-coolant conditions. The number of experiments conducted permitted determination of the range of tube ruptures for which high peak cladding temperatures could result in the Semiscale Mod-1 system. The paper contains a description of the Semiscale Mod-1 system and a discussion of the steam generator tube rupture tests conducted. The experimental results from the test series and the thermal-hydraulic phenomena found to influence the core thermal response during the experiments are discussed

  13. Common and uncommon CT findings in rupture and impending rupture of abdominal aortic aneurysms

    International Nuclear Information System (INIS)

    Ahmed, M.Z.; Ling, L.; Ettles, D.F.

    2013-01-01

    The rapid imaging evaluation and diagnosis of rupture and impending rupture of an abdominal aortic aneurysm (AAA) is imperative. This article describes the imaging findings of rupture, impending rupture, and other abdominal aortic abnormalities. It is important not to overlook AAA as the consequences can be life threatening. All patients who had open or endovascular repair of AAA rupture over 6 years (2008–2012) were identified from our departmental database. The computed tomography (CT) images of 99 patients were reviewed for relevant findings. The mean age of the patients was 65 years and 85% were male

  14. Radionuclide scan findings in delayed splenic rupture

    International Nuclear Information System (INIS)

    Flickinger, F.W.; Jackson, G.L.

    1978-01-01

    An initial liver/spleen scan performed on a patient with blunt abdominal trauma was negative 3 days following the accident. A follow-up scan 7 days later showed definite evidence of splenic rupture, proved surgically. The authors conclude that, in such cases, spleen scans may be negative initially because of delayed splenic ruptures

  15. Rupture of esophagus by compressed air.

    Science.gov (United States)

    Wu, Jie; Tan, Yuyong; Huo, Jirong

    2016-11-01

    Currently, beverages containing compressed air such as cola and champagne are widely used in our daily life. Improper ways to unscrew the bottle, usually by teeth, could lead to an injury, even a rupture of the esophagus. This letter to editor describes a case of esophageal rupture caused by compressed air.

  16. Coiling of ruptured pericallosal artery aneurysms.

    NARCIS (Netherlands)

    Menovsky, T.; Rooij, W.J.J. van; Sluzewski, M.; Wijnalda, D.

    2002-01-01

    OBJECTIVE: To assess the technical feasibility of treating ruptured pericallosal artery aneurysms with detachable coils and to evaluate the anatomic and clinical results. METHODS: Over a period of 27 months, 12 patients with a ruptured pericallosal artery aneurysm were treated with detachable

  17. Treatment Alternative for Irreparable Rotator Cuff Ruptures ...

    African Journals Online (AJOL)

    2016-09-03

    Sep 3, 2016 ... and such ruptures also lead to a pseudo-paralysis.[1,2]. Pain during daily ... 2) repairable rotator cuff rupture, as determined on MRI and during arthroscopy ..... functioning and lead to cosmetic deformities.[4]. Arthroplasty is a ...

  18. Uterine rupture without previous caesarean delivery

    DEFF Research Database (Denmark)

    Thisted, Dorthe L. A.; H. Mortensen, Laust; Krebs, Lone

    2015-01-01

    to uterine rupture when adjusted for parity, epidural analgesia and augmentation by oxytocin. CONCLUSION: Although uterine rupture is rare, its association with epidural analgesia and augmentation of labour with oxytocin in multipara should be considered. Thus, vigilance should be exercised when labour...

  19. Spontaneous rupture of choledochal cyst: case report

    International Nuclear Information System (INIS)

    Shin, Ho Seob; Nam, Kyung Jin; Lee, Jin Hwa; Kim, Chan Sung; Choi, Jong Cheol; Oh, Jong Young

    2002-01-01

    Spontaneous rupture of a choledochal cyst leading to biliary peritonitis is a rare complication which can be fatal if not promptly diagnosed. The authors report the ultrasound and CT findings of two cases of spontaneous choledochal cystic rupture and the biliary peritonitis which ensued

  20. Isolated gallbladder rupture following blunt abdominal injury

    African Journals Online (AJOL)

    2015-05-26

    May 26, 2015 ... Kaohsiung Medical University, Kaohsiung, Taiwan. Abstract. Isolated traumatic gallbladder rupture subsequent to blunt abdominal injury is rare. Most literatures on the subjects consist of case reports. We reported a rare case of isolated gallbladder rupture and discussed the possible predisposing factors to ...

  1. Spontaneous rupture of choledochal cyst: case report

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Seob; Nam, Kyung Jin; Lee, Jin Hwa; Kim, Chan Sung; Choi, Jong Cheol; Oh, Jong Young [Dong-a University College of Medicine, Pusan (Korea, Republic of)

    2002-11-01

    Spontaneous rupture of a choledochal cyst leading to biliary peritonitis is a rare complication which can be fatal if not promptly diagnosed. The authors report the ultrasound and CT findings of two cases of spontaneous choledochal cystic rupture and the biliary peritonitis which ensued.

  2. Protection Performance Simulation of Coal Tar-Coated Pipes Buried in a Domestic Nuclear Power Plant Using Cathodic Protection and FEM Method

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Y.; Lim, B. T.; Kim, K. S.; Kim, J. W.; Park, H. B. [KEPCO Engineering and Construction Company, Gimcheon (Korea, Republic of); Kim, Y. S.; Kim, K. T. [Andong National University, Andong (Korea, Republic of)

    2017-06-15

    Coal tar-coated pipes buried in a domestic nuclear power plant have operated under the cathodic protection. This work conducted the simulation of the coating performance of these pipes using a FEM method. The pipes, being ductile cast iron have been suffered under considerably high cathodic protection condition beyond the appropriate condition. However, cathodic potential measured at the site revealed non-protected status. Converting from 3D CAD data of the power plant to appropriate type for a FEM simulation was conducted and cathodic potential under the applied voltage and current was calculated using primary and secondary current distribution and physical conditions. FEM simulation for coal tar-coated pipe without defects revealed over-protection condition if the pipes were well-coated. However, the simulation for coal tar-coated pipes with many defects predict that the coated pipes may be severely degraded. Therefore, for high risk pipes, direct examination and repair or renewal of pipes are strongly recommended.

  3. Trend of field data on pipe wall thinning for BWR power plants

    International Nuclear Information System (INIS)

    Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

    2009-01-01

    Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

  4. 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)

  5. 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.)

  6. A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, Robert A.

    2013-09-18

    Hydrogen explosions may occur simultaneously with water hammer accidents in nuclear facilities, and a theoretical mechanism to relate water hammer to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in pipe systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the pipe system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany. Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism.

  7. MRI of tibialis anterior tendon rupture

    International Nuclear Information System (INIS)

    Gallo, Robert A.; DeMeo, Patrick J.; Kolman, Brett H.; Daffner, Richard H.; Sciulli, Robert L.; Roberts, Catherine C.

    2004-01-01

    Ruptures of the tibialis anterior tendon are rare. We present the clinical histories and MRI findings of three recent male patients with tibialis anterior tendon rupture aged 58-67 years, all of whom presented with pain over the dorsum of the ankle. Two of the three patients presented with complete rupture showing discontinuity of the tendon, thickening of the retracted portion of the tendon, and excess fluid in the tendon sheath. One patient demonstrated a partial tear showing an attenuated tendon with increased surrounding fluid. Although rupture of the tibialis anterior tendon is a rarely reported entity, MRI is a useful modality in the definitive detection and characterization of tibialis anterior tendon ruptures. (orig.)

  8. Ruptured gastroepiploic artery aneurysm: A case report

    Directory of Open Access Journals (Sweden)

    Ahmad S. Ashrafi

    Full Text Available Introduction: Gastroepiploic artery aneurysms are extremely rare, with few reported cases in the literature. The risk of rupture however, is high and thus warrants attention. Presentation of case: Here we present a rare case of a women who presented to the emergency department in shock and was found to have a ruptured gastroepiploic artery aneurysm during surgical exploration. Suture ligation of the aneurysm was completed. Discussion: Although rare, gastroepiploic artery aneurysms have up to a 90% rate of rupture and therefore require intervention. A laparoscopic approach has been described however, in cases where rupture has occurred, urgent laparotomy and control of hemorrhage is needed. Conclusion: We describe a rare case of a ruptured gastroepiploic aneurysm that was successfully managed with urgent laparotomy and aneurysmal resection. Keywords: Gastroepiploic, Aneurysm, Hemorrhage, Case report

  9. Waste pipe calculus extensions

    International Nuclear Information System (INIS)

    O'Connell, W.J.

    1979-01-01

    The waste pipe calculus provides a rapid method, using Laplace transforms, to calculate the transport of a pollutant such as nuclear waste, by a network of one-dimensional flow paths. The present note extends previous work as follows: (1) It provides an alternate approximation to the time-domain function (inverse Laplace transform) for the resulting transport. This algebraic approximation may be viewed as a simpler and more approximate model of the transport process. (2) It identifies two scalar quantities which may be used as summary consequence measures of the waste transport (or inversely, waste retention) system, and provides algebraic expressions for them. (3) It includes the effects of radioactive decay on the scalar quantity results, and further provides simplifying approximations for the cases of medium and long half-lives. This algebraic method can be used for quick approximate analyses of expected results, uncertainty and sensitivity, in evaluating selection and design choices for nuclear waste disposal systems

  10. Solar chemical heat pipe

    International Nuclear Information System (INIS)

    Levy, M.; Levitan, R.; Rosin, H.; Rubin, R.

    1991-08-01

    The performance of a solar chemical heat pipe was studied using CO 2 reforming of methane as a vehicle for storage and transport of solar energy. The endothermic reforming reaction was carried out in an Inconel reactor, packed with a Rh catalyst. The reactor was suspended in an insulated box receiver which was placed in the focal plane of the Schaeffer Solar Furnace of the Weizman Institute of Science. The exothermic methanation reaction was run in a 6-stage adiabatic reactor filled with the same Rh catalyst. Conversions of over 80% were achieved for both reactions. In the closed loop mode the products from the reformer and from the metanator were compressed into separate storage tanks. The two reactions were run either separately or 'on-line'. The complete process was repeated for over 60 cycles. The overall performance of the closed loop was quite satisfactory and scale-up work is in progress in the Solar Tower. (authors). 35 refs., 2 figs

  11. Subsea pipe dream

    Energy Technology Data Exchange (ETDEWEB)

    Balcombe, Mark

    1988-09-22

    The Gulf of Mexico is famous today mainly for the ferocity of its hurricanes. But for anyone in the oil industry, it is also known for the vast array of oil pipelines that criss-cross its stormy waters, and for the large number of pipeline-laying barges which install them. Soon many of these vessels could be steaming to British waters - not to escape the weather, but to cash in on a bonanza of pipe-laying activity which could soon take place offshore northern Europe. The construction of new pipelines off the UK, Norway and Netherlands will, however, present a new range of challenges for pipeline designers and builders. First and foremost is the Piper Alpha platform disaster, which could saddle the UK offshore industry with a Pound 500 million-plus bill for the installation of emergency shutdown valves (ESVs) on existing lines.

  12. Steam line rupture experiments with the PPOOLEX test facility

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2008-07-01

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  13. Steam line rupture experiments with the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  14. Monitoring of pipe displacements in French LMFBR SUPERPHENIX

    International Nuclear Information System (INIS)

    Foucher, N.; Debaene, J.P.; Renault, Y.; Blin, B.

    1993-01-01

    In order to check that pipe supports work properly and that the locking of snubbers or the loss of supports do not put a pipe in unacceptable loading conditions, a monitoring of the behaviour of the main pipes of SUPERPHENIX is planned. This monitoring system consists in measuring the displacements at selected points of the pipe by means of measuring rods and checking that these displacements remain inside allowable domains. These allowable domains are defined so that, if the displacements of the pipe are inside all these domains, the plant operator is sure that the stresses verify the allowable limits and then no additional inspection is carried out. In the opposite case, the operator will inspect the pipe in detail in order to determine the consequences and repair if necessary before restarting. Selection of points for monitoring was done with the to minimize the number of measures to be carried out and to use as far as possible the measuring rods that were installed to check that pipe displacements were consistent with what has been obtained in design calculations. However, it appears necessary to ensure that any incident occurring at any point of the pipe can be detected and, if necessary, additional measuring rods may be installed. An incident is said detectable if it induces on at least one measuring rod a deviation with respect to expected displacement not lower than 5 mm. It has been chosen so that small normal changes in measured displacements are not mistaken as incidents. The incidents that are supposed likely to occur are: 1) loss of a support which induces mainly primary stresses, 2) locking of a snubber which induces mainly secondary stresses. Monitoring of pipe displacements is a simple and effective way of checking that no damaging perturbation has occurred on the pipe. Calculations carried out on the DHR loops of SUPERPHENIX show that allowable domains of acceptable size may be obtained using a relatively small number of measuring rods. The method

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Earthquake lights and rupture processes

    Directory of Open Access Journals (Sweden)

    T. V. Losseva

    2005-01-01

    Full Text Available A physical model of earthquake lights is proposed. It is suggested that the magnetic diffusion from the electric and magnetic fields source region is a dominant process, explaining rather high localization of the light flashes. A 3D numerical code allowing to take into account the arbitrary distribution of currents caused by ground motion, conductivity in the ground and at its surface, including the existence of sea water above the epicenter or (and near the ruptured segments of the fault have been developed. Simulations for the 1995 Kobe earthquake were conducted taking into account the existence of sea water with realistic geometry of shores. The results do not contradict the eyewitness reports and scarce measurements of the electric and magnetic fields at large distances from the epicenter.

  17. Traumatic rupture of the aorta

    International Nuclear Information System (INIS)

    Dorfman, G.S.; Paolella, L.P.; Haas, R.A.; Lambiase, R.E.; Cronan, J.J.

    1988-01-01

    To evaluate the acceptability of digital subtraction angiography (DSA) in the evaluation of traumatic rupture of the aorta (TRA), the authors obtained 56 thoracic aortograms in 55 consecutive trauma patients, using both DSA and cut-film angiography (CFA). Both studies were ranked blindly and assigned scores for quality and diagnosis. Interobserver variance for DSA and CFA quality rankings was insignificant. CFA achieved significantly higher ranking for quality. In the 56 examinations, CFA demonstrated seven abnormalities that demanded intervention for follow-up angiography. DSA demonstrated only five of these and found no additional abnormalities. While this evaluation applies only to the particular digital system that the authors tested, they found that DSA, as compared with CFA, failed to demonstrate significant aortic injury. The confidence of diagnosis was significantly greater with CFA. Similar double-blind evaluation is mandatory at any trauma center prior to converting from CFA to DSA in the diagnosis of this life-threatening condition

  18. Rupture of the meniscofibular ligament

    Directory of Open Access Journals (Sweden)

    Poyanli Oguz

    2010-05-01

    Full Text Available Abstract The meniscofibular ligament is an anatomically defined ligament of the knee in humans. However, there are no data regarding the prognosis following injury to this ligament. Our case was a 42-year-old man who presented at our clinic with pain of the lateral side of his left knee. MRI of his left knee revealed the rupture of the meniscofibular ligament. The mechanism of injury was consistent with anatomical and mechanical studies of the meniscofibular ligament. The patient was treated conservatively for 1 year, but his pain did not resolve completely. A case series of patients with the same injury is required to establish an effective treatment for this rare injury.

  19. Role of theoretical dynamics in vibration diagnostics of pipe systems

    International Nuclear Information System (INIS)

    Rejent, B.

    1992-01-01

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

  20. Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

    2009-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

  1. Superconducting pipes and levitating magnets.

    Science.gov (United States)

    Levin, Yan; Rizzato, Felipe B

    2006-12-01

    Motivated by a beautiful demonstration of the Faraday and the Lenz laws in which a small neodymium magnet falls slowly through a conducting nonferromagnetic tube, we consider the dynamics of a magnet falling coaxially through a superconducting pipe. Unlike the case of normal conducting pipes, in which the magnet quickly reaches the terminal velocity, inside a superconducting tube the magnet falls freely. On the other hand, to enter the pipe the magnet must overcome a large electromagnetic energy barrier. For sufficiently strong magnets, the barrier is so large that the magnet will not be able to penetrate it and will be levitated over the mouth of the pipe. We calculate the work that must done to force the magnet to enter a superconducting tube. The calculations show that superconducting pipes are very efficient at screening magnetic fields. For example, the magnetic field of a dipole at the center of a short pipe of radius a and length L approximately > a decays, in the axial direction, with a characteristic length xi approximately 0.26a. The efficient screening of the magnetic field might be useful for shielding highly sensitive superconducting quantum interference devices. Finally, the motion of the magnet through a superconducting pipe is compared and contrasted to the flow of ions through a trans-membrane channel.

  2. In service inspection of pipes based on risk methods

    International Nuclear Information System (INIS)

    Mendoza G, G.; Viais J, J.; Carmona C, M.

    2006-01-01

    The politics of the Nuclear Regulatory Commission (by its initials in English NRC) of the United States of America on the use of the Probabilistic Safety Analysis (PSA) in activities of nuclear regulation it foments the use of this analysis technique to improve the decisions making, to reduce the unnecessary work in maintenance aspects, inspection and tests and to improve the regulatory efficiency. The inspection programs in service (ISI by its initials in English) developed by the American Society of Mechanical Engineers (by its initials in English ASME) it has been the one primary mechanism to prove the mechanical equipment in plants of nuclear energy, these programs indeed have been carried out in plants of nuclear energy by more of two decades. Their purpose is to identify the conditions, such as indications of cracks that are precursory of flights and ruptures which violate the integrity principles of the pressure frontier. The inspection in service activities include ultrasonic tests, surface tests and penetrating liquids test, also activities that include the scaffolds construction, removal of insulations and welding polishing. The inspections in service every 18 months during the times outside of service are executed. One of the objectives is to lower the costs of the inspections during the times outside of service and to reduce the exposure to the radiation by part of the personnel during these times out for inspections, while it is increased or it maintains the personnel's safety and the reliability. As part of the methodology a pipe segment is selected for which a fault in any point has the same consequences, being calculated the fault probability of the tube using the dimensions of the segment. In this work the inspection in service methodology is applied based on risk to an aspersion system of low pressure of the Laguna Verde Nucleo electric Central. For this system a reduction in the number of welding to inspect of 103 to only 15 is obtained

  3. WWER-1000/320 steam generator collector rupture. Radiological consequences

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, A; Sartmadzhiev, A; Balabanov, E [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    A model describing a hypothetical accident with direct release of primary coolant to the atmosphere is proposed. Cover lifting of the primary collector due to a rupture of the fixing bolts leads to a coolant release. The initial and boundary conditions of the accident scenario have been selected to provide for the most unfavorable conditions. The total release of primary coolant during the first 15 min of transient are estimated to 50.8 tons, of these 48.5 t with the initial activity in the primary coolant circuit. Without evacuation or sheltering, after 7 days of exposure, the expected dose at the boundary of the restricted zone is 0.0182 Sv for the whole body and 0.184 Sv for the thyroid gland. The effective equivalent dose on the site would be 0.0521 Sv. As a result of the analysis it is concluded that the steam generator collector rupture is not jeopardizing the core heat removal even with a minimum configuration of ECCS as the cooling is accomplished through the steam generators. The radiological consequences of the accident would be relatively small if an emergency procedure is applied at the 15-th minute of the transient. 1 ref.

  4. Loop Heat Pipe Startup Behaviors

    Science.gov (United States)

    Ku, Jentung

    2016-01-01

    A loop heat pipe must start successfully before it can commence its service. The startup transient represents one of the most complex phenomena in the loop heat pipe operation. This paper discusses various aspects of loop heat pipe startup behaviors. Topics include the four startup scenarios, the initial fluid distribution between the evaporator and reservoir that determines the startup scenario, factors that affect the fluid distribution between the evaporator and reservoir, difficulties encountered during the low power startup, and methods to enhance the startup success. Also addressed are the pressure spike and pressure surge during the startup transient, and repeated cycles of loop startup and shutdown under certain conditions.

  5. Research notes : drainage facility asset management : more than an inventory of pipes.

    Science.gov (United States)

    2007-04-01

    The primary objectives for the research project were twofold: 1) To develop and implement an Oregon-specific system for inventorying and evaluating the condition of pipes, culverts, and stormwater facilities based on the FHWA Culvert Management Syste...

  6. Leachate storage transport tanker loadout piping

    International Nuclear Information System (INIS)

    Whitlock, R.W.

    1994-01-01

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

  7. Determination of the pipe stemming load

    International Nuclear Information System (INIS)

    Cowin, S.C.

    1979-01-01

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

  8. Material Parameters for Creep Rupture of Austenitic Stainless Steel Foils

    Science.gov (United States)

    Osman, H.; Borhana, A.; Tamin, M. N.

    2014-08-01

    Creep rupture properties of austenitic stainless steel foil, 347SS, used in compact recuperators have been evaluated at 700 °C in the stress range of 54-221 MPa to establish the baseline behavior for its extended use. Creep curves of the foil show that the primary creep stage is brief and creep life is dominated by tertiary creep deformation with rupture lives in the range of 10-2000 h. Results are compared with properties of bulk specimens tested at 98 and 162 MPa. Thin foil 347SS specimens were found to have higher creep rates and higher rupture ductility than their bulk specimen counterparts. Power law relationship was obtained between the minimum creep rate and the applied stress with stress exponent value, n = 5.7. The value of the stress exponent is indicative of the rate-controlling deformation mechanism associated with dislocation creep. Nucleation of voids mainly occurred at second-phase particles (chromium-rich M23C6 carbides) that are present in the metal matrix by decohesion of the particle-matrix interface. The improvement in strength is attributed to the precipitation of fine niobium carbides in the matrix that act as obstacles to the movement of dislocations.

  9. Metrics for comparing dynamic earthquake rupture simulations

    Science.gov (United States)

    Barall, Michael; Harris, Ruth A.

    2014-01-01

    Earthquakes are complex events that involve a myriad of interactions among multiple geologic features and processes. One of the tools that is available to assist with their study is computer simulation, particularly dynamic rupture simulation. A dynamic rupture simulation is a numerical model of the physical processes that occur during an earthquake. Starting with the fault geometry, friction constitutive law, initial stress conditions, and assumptions about the condition and response of the near‐fault rocks, a dynamic earthquake rupture simulation calculates the evolution of fault slip and stress over time as part of the elastodynamic numerical solution (Ⓔ see the simulation description in the electronic supplement to this article). The complexity of the computations in a dynamic rupture simulation make it challenging to verify that the computer code is operating as intended, because there are no exact analytic solutions against which these codes’ results can be directly compared. One approach for checking if dynamic rupture computer codes are working satisfactorily is to compare each code’s results with the results of other dynamic rupture codes running the same earthquake simulation benchmark. To perform such a comparison consistently, it is necessary to have quantitative metrics. In this paper, we present a new method for quantitatively comparing the results of dynamic earthquake rupture computer simulation codes.

  10. Hepatic Rupture Induced by Spontaneous Intrahepatic Hematoma

    Directory of Open Access Journals (Sweden)

    Jin-bao Zhou

    2018-01-01

    Full Text Available The etiology of hepatic rupture is usually secondary to trauma, and hepatic rupture induced by spontaneous intrahepatic hematoma is clinically rare. We describe here a 61-year-old female patient who was transferred to our hospital with hepatic rupture induced by spontaneous intrahepatic hematoma. The patient had no history of trauma and had a history of systemic lupus erythematosus for five years, taking a daily dose of 5 mg prednisone for treatment. The patients experienced durative blunt acute right upper abdominal pain one day after satiation, which aggravated in two hours, accompanied by dizziness and sweating. Preoperative diagnosis was rupture of the liver mass. Laparotomy revealed 2500 mL fluid consisting of a mixture of blood and clot in the peritoneal cavity. A 3.5 cm × 2.5 cm rupture was discovered on the hepatic caudate lobe near the vena cava with active arterial bleeding, and a 5  × 6 cm hematoma was reached on the right posterior lobe of the liver. Abdominal computed tomography (CT and laparotomy revealed spontaneous rupture of intrahepatic hematoma with hemorrhagic shock. The patient was successfully managed by suturing the rupture of the hepatic caudate lobe and clearing part of the hematoma. The postoperative course was uneventful, and the patient was discharged after two weeks of hospitalization.

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

    International Nuclear Information System (INIS)

    Zhong Zhimin; Li Jinsong; Zheng Hui

    2012-01-01

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

  12. 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

  13. Pipe restraints for nuclear power plants

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  14. Does Full Wound Rupture following Median Pilonidal Closure Alter Long-Term Recurrence Rate?

    Science.gov (United States)

    Doll, Dietrich; Matevossian, Edouard; Luedi, Markus M; Schneider, Ralf; van Zypen, Dominic; Novotny, Alexander

    2015-01-01

    The purpose of this study was to examine the recurrence rate of wound rupture in primary pilonidal sinus disease (PSD) after median closure. A total of 583 patients from the German military cohort were interviewed. We compared the choice of surgical therapy, wound dehiscence (if present) and long-term recurrence-free survival for patients with primary open treatment, marsupialization and primary median treatment (closed vs. secondary open, respectively). Actuarial recurrence rate was determined using the Kaplan-Meier calculation with a follow-up of up to 20 years after primary PSD surgery. Patients with excision followed by primary open wound treatment showed a significantly lower 5- than 10-year recurrence rate (8.3 vs. 11.2%) compared to the patients with primary midline closure (17.4 vs. 20.5%, p = 0.03). The 20-year recurrence rate was 28% in primary open wound treatment versus 44% in primary midline closure without wound rupture. In contrast to these findings, long-term recurrence rates following secondary open wound treatment (12.2% at 5 years vs. 17.1% at 10 years) tended to be higher (although not significantly, p = 0.57) compared to primary open treatment (8.3% at 5 years vs. 11.2% at 10 years). There was no statistical difference in long-term recurrence rates between secondary open and primary midline closure (p = 0.7). Hence, despite only a short wound closure time experienced before wound rupture, the patient does not fully benefit from an open wound treatment in terms of recurrence rate. The postoperative pilonidal sinus wound rupture of primary midline closures did not significantly increase the 5- and 10-year long-term recurrence rates compared to uneventfully healing primary midline closures. © 2015 S. Karger AG, Basel.

  15. Effect of sodium environment on the creep-rupture and low-cycle fatigue behavior of austenitic stainless steels

    International Nuclear Information System (INIS)

    Natesan, K.; Chopra, D.K.; Zeman, G.J.; Smith, D.L.; Kassner, T.F.

    1977-01-01

    Austenitic stainless steels used for in-core structural components, piping, valves, and the intermediate heat exchanger in Liquid-Metal Fast-Breeder Reactors (LMFBRs) are subjected to sodium at elevated temperatures and to complex stress conditions. As a result, the materials can undergo compositional and microstructural changes as well as mechanical deformation by creep and cyclic fatigue processes. In the present paper, information is presented on the creep-rupture and low-cycle fatigue behavior of Types 304 and 316 stainless steel in the solution-annealed condition and after long-term exposure to flowing sodium. The nonmetallic impurity-element concentrations in the sodium were controlled at levels similar to those in EBR-II primary sodium. Strain-time relationships developed from the experimental creep data were used to generate isochronous stress-creep strain curves as functions of sodium-exposure time and temperature. The low-cycle fatigue data were used to obtain relationships between plastic strain range and cycles-to-failure based on the Coffin-Manson formalism and a damage-rate approach developed at ANL. An analysis of the cyclic stress-strain behavior of the materials showed that the strain-hardening rates for the sodium-exposed steels were larger than those for the annealed material. However, the sodium-exposed specimens showed significant softening, as evidenced by the lower stress at half the fatigue life. Microstructural information obtained from the different specimens suggests that crack initiation is more difficult in the long-term sodium-exposed specimens when compared with the solution-annealed material. Based on the expected carbon concentrations in LMFBR primary system sodium, moderate carburization of the austenitic stainless steels will not degrade the mechanical properties to a significant extent, and therefore, will not limit the performance of out-of-core components. (author)

  16. Treatment of traumatic rupture of the thoracic aorta

    Directory of Open Access Journals (Sweden)

    Davidović Lazar

    2008-01-01

    Full Text Available INTRODUCTION Interest for traumatic thoracic aorta rupture stems from the fact that its number continually increases, and it can be rapidly lethal. OBJECTIVE The aim of this study is to present early and long term results as well as experiences of our team in surgical treatment of traumatic thoracic aorta rupture. METHOD Our retrospective study includes 12 patients with traumatic thoracic aorta rupture treated between 1985 and 2007. There were 10 male and two female patients of average age 30.75 years (18-74. RESULTS In six cases, primary diagnosis was established during the first seven days days after trauma, while in 6 more than one month later. In 11 cases, classical open surgical procedure was performed, while endovascular treatment was used in one patient. Three (25% patients died, while two (16.6% had paraplegia. Nine patients (75% were treated without complications, and are in good condition after a mean follow-up period of 9.7 years (from one month to 22 years. CONCLUSION Surgical treatment requires spinal cord protection to prevent paraplegia, using cardiopulmonary by-pass (three of our cases or external heparin-bonded shunts (five of our cases. Cardiopulmonary by-pass is followed with lower incidence of paraplegia, however it is not such a good solution for patients with polytrauma because of hemorrhage. The endovascular repair is a safe and feasible procedure in the acute phase, especially because of traumatic shock and polytrauma which contributes to higher mortality rate after open surgery. On the other hand, in chronic postrauamatic aortic rupture, open surgical treatment is connected with a lower mortality rate and good long-term results. There have been no published data about long-term results of endovascular treatment in the chronic phase.

  17. Misdiagnosed Chest Pain: Spontaneous Esophageal Rupture

    Science.gov (United States)

    Inci, Sinan; Gundogdu, Fuat; Gungor, Hasan; Arslan, Sakir; Turkyilmaz, Atila; Eroglu, Atila

    2013-01-01

    Chest pain is one of themost common complaints expressed by patients presenting to the emergency department, and any initial evaluation should always consider life-threatening causes. Esophageal rupture is a serious condition with a highmortality rate. If diagnosed, successful therapy depends on the size of the rupture and the time elapsed between rupture and diagnosis.We report on a 41-year-old woman who presented to the emergency department complaining of left-sided chest pain for two hours. PMID:27122690

  18. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Schwab, P.R.

    1981-01-01

    Creep rupture is a long term failure mode in structural materials that occurs at high temperatures and moderate stress levels. The deterioration of the material preceding rupture, termed creep damage, manifests itself in the formation of small cavities on grain boundaries. To measure creep damage, sometimes uniaxial tests are performed, sometimes density measurements are made, and sometimes the grain boundary cavities are measured by microscopy techniques. The purpose of the present research is to explore a new method of measuring creep rupture, which involves measuring the curvature of eccentric pressurized tubes. Theoretical investigations as well as the design, construction, and operation of an experimental apparatus are included in this research

  19. Spontaneous distal rupture of the plantar fascia.

    Science.gov (United States)

    Gitto, Salvatore; Draghi, Ferdinando

    2018-07-01

    Spontaneous ruptures of the plantar fascia are uncommon injuries. They typically occur at its calcaneal insertion and usually represent a complication of plantar fasciitis and local treatment with steroid injections. In contrast, distal ruptures commonly result from traumatic injuries. We describe the case of a spontaneous distal rupture of the plantar fascia in a 48-year-old woman with a low level of physical activity and no history of direct injury to the foot, plantar fasciitis, or steroid injections. © 2017 Wiley Periodicals, Inc.

  20. 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

  1. Principal working group 3 on primary circuit integrity

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-31

    The main themes of this conference (13 papers) are: operating experience on leakages and failures in nuclear power plant piping, coolant circuits and steam generator tubes, probabilistic estimation and risk assessment, system failure analysis, leakage events and frequency, leak rate models and crack propagation mechanics, damage mechanisms and rupture probability.

  2. Principal working group 3 on primary circuit integrity

    International Nuclear Information System (INIS)

    1992-01-01

    The main themes of this conference (13 papers) are: operating experience on leakages and failures in nuclear power plant piping, coolant circuits and steam generator tubes, probabilistic estimation and risk assessment, system failure analysis, leakage events and frequency, leak rate models and crack propagation mechanics, damage mechanisms and rupture probability

  3. Heat pipe turbine vane cooling

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

  4. Effects of swirl in turbulent pipe flows : computational studies

    Energy Technology Data Exchange (ETDEWEB)

    Nygaard, Frode

    2011-07-01

    The primary objective of this doctoral thesis was to investigate the effect of swirl in steady turbulent pipe flows. The work has been carried out by a numerical approach, with direct numerical simulations as the method of choice. A key target to pursue was the effects of the swirl on the wall friction in turbulent pipe flows. The motivation came from studies of rotating pipe flows in which drag reduction was achieved. Drag reduction was reported to be due to the swirl favourably influencing the coherent turbulent structures in the near-wall region. Based on this, it was decided to investigate if similar behaviour could be obtained by inducing a swirl in a pipe with a stationary wall. To do a thorough investigation of the general three-dimensional swirl flow and particularly of the swirl effects; chosen variations of mean and turbulent flow parameters were explored together with complementary flow visualizations. Two different approaches in order to induce the swirl in the turbulent pipe flow, have been carried out. However, the present thesis might be regarded to be comprised of three parts. The first part consists of the first approach to induce the swirl. Here a prescribed circumferential force was implemented in a serial open source Navier-Stokes solver. In the second approach, the swirl was intended induced by implementing structures at the wall. Simulations of flows through a pipe with one or more helical fin(s) at the pipe wall was decided to be performed. In order to pursue this approach, it was found necessary to do a parallelization of the existing serial numerical code. The key element of this parallelization has been included as a part of the present work. Additionally, the helical fin(s) were implemented into the code by use of an immersed boundary method. A validation of this work is also documented in the thesis. The work done by parallelizing the code and implementing an immersed boundary method constitutes the second part of the present thesis. The

  5. Reliability analysis of stiff versus flexible piping

    International Nuclear Information System (INIS)

    Lu, S.C.

    1985-01-01

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

  6. Piping inspection round robin

    International Nuclear Information System (INIS)

    Heasler, P.G.; Doctor, S.R.

    1996-04-01

    The piping inspection round robin was conducted in 1981 at the Pacific Northwest National Laboratory (PNNL) to quantify the capability of ultrasonics for inservice inspection and to address some aspects of reliability for this type of nondestructive evaluation (NDE). The round robin measured the crack detection capabilities of seven field inspection teams who employed procedures that met or exceeded the 1977 edition through the 1978 addenda of the American Society of Mechanical Engineers (ASME) Section 11 Code requirements. Three different types of materials were employed in the study (cast stainless steel, clad ferritic, and wrought stainless steel), and two different types of flaws were implanted into the specimens (intergranular stress corrosion cracks (IGSCCs) and thermal fatigue cracks (TFCs)). When considering near-side inspection, far-side inspection, and false call rate, the overall performance was found to be best in clad ferritic, less effective in wrought stainless steel and the worst in cast stainless steel. Depth sizing performance showed little correlation with the true crack depths

  7. 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

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

    International Nuclear Information System (INIS)

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

    1995-04-01

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

  9. Real-Time Detection of Rupture Development: Earthquake Early Warning Using P Waves From Growing Ruptures

    Science.gov (United States)

    Kodera, Yuki

    2018-01-01

    Large earthquakes with long rupture durations emit P wave energy throughout the rupture period. Incorporating late-onset P waves into earthquake early warning (EEW) algorithms could contribute to robust predictions of strong ground motion. Here I describe a technique to detect in real time P waves from growing ruptures to improve the timeliness of an EEW algorithm based on seismic wavefield estimation. The proposed P wave detector, which employs a simple polarization analysis, successfully detected P waves from strong motion generation areas of the 2011 Mw 9.0 Tohoku-oki earthquake rupture. An analysis using 23 large (M ≥ 7) events from Japan confirmed that seismic intensity predictions based on the P wave detector significantly increased lead times without appreciably decreasing the prediction accuracy. P waves from growing ruptures, being one of the fastest carriers of information on ongoing rupture development, have the potential to improve the performance of EEW systems.

  10. Fatigue crack growth in austenitic stainless steel piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Cheissoux, J.L.; Lebey, J.

    1981-04-01

    The study presented in this paper is being carried out with a view to substantiating the calculations of the fatigue crack growth in pipes made of 316 L stainless steel. The results obtained may be applied to P.W.R. primary piping. It is divided into two parts. First, fatigue tests (cyclic pressure) are carried out under hot and cold conditions with straight pipes machined with notches of various dimensions. The crack propagation and the fatigue crack growth rate are measured here. Second, calculations are made in order to interpret experimental results. From elastic calculations the stress intensity factor is assessed to predict the crack growth rate. The results obtained until now and presented in this paper relate to longitudinal notches

  11. The application of viscous dampers as pipe restraints

    International Nuclear Information System (INIS)

    Keowen, R.S.; Hueffmann, G.; Mays, B.; Rencher, D.

    1993-01-01

    Dynamic loading of power generation piping systems may result in nonpermissable deflections and stresses. Fatigue failure translate to increased maintenance costs and possible lost revenue. Undesirable loading can occur due to external events such as earthquakes and internal events such as water and steam hammer, two-phase flow and cavitation. Sway braces and snubbers have been employed to reduce the negative effects of piping motion in emergency cases, however, repetitive loading due to internal events has caused premature wear and failure. Visco elastic dampers, however, have proven to be piping response due to slugging, steam hammer and other repetitive loads. Functional and modeling aspects of visco elastic dampers are discussed, experimental evidence of their effectiveness in a steam hammer application is presented and examples of primary coolant loop restraint applications are illustrated

  12. Geological and Petrographic Characteristics of Kimberlite Pipes

    Directory of Open Access Journals (Sweden)

    N. N. Zinchuk

    2016-12-01

    Full Text Available Studies of the geological structure and petrochemical composition of the Siberian Platform kimberlites indicated complexity, diversity of geological, tectonic, and paleogeographic situations, which must be considered for proper prospecting-exploration for diamonds in each area of investigation. Information about petrochemical composition of potential diatremes, hosting, and overlying sedimentary and magmatic formations is an important prerequisite for prospecting of kimberlite deposits in different geologic-tectonic conditions. The most attention should be paid to typomorphic specific features of primary and secondary minerals of diatremes. Each diamondiferous region is characterized by a certain set of typomorphic associations of kimberlites primary and secondary minerals. The diamonds with ultrabasic association of solid phase inclusions (olivine, chrome-spinel, pyrope, etc. dominate in majority of kimberlite pipes.

  13. Acoustic leak detection in piping systems, 4

    International Nuclear Information System (INIS)

    Kitajima, Akira; Naohara, Nobuyuki; Aihara, Akihiko

    1983-01-01

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

  14. Finding buried metallic pipes using a non-destructive approach based on 3D time-domain induced polarization data

    Science.gov (United States)

    Shao, Zhenlu; Revil, André; Mao, Deqiang; Wang, Deming

    2018-04-01

    The location of buried utility pipes is often unknown. We use the time-domain induced polarization method to non-intrusively localize metallic pipes. A new approach, based on injecting a primary electrical current between a pair of electrodes and measuring the time-lapse voltage response on a set of potential electrodes after shutting down this primary current is used. The secondary voltage is measured on all the electrodes with respect to a single electrode used as a reference for the electrical potential, in a way similar to a self-potential time lapse survey. This secondary voltage is due to the formation of a secondary current density in the ground associated with the polarization of the metallic pipes. An algorithm is designed to localize the metallic object using the secondary voltage distribution by performing a tomography of the secondary source current density associated with the polarization of the pipes. This algorithm is first benchmarked on a synthetic case. Then, two laboratory sandbox experiments are performed with buried metallic pipes located in a sandbox filled with some clean sand. In Experiment #1, we use a horizontal copper pipe while in Experiment #2 we use an inclined stainless steel pipe. The result shows that the method is effective in localizing these two pipes. At the opposite, electrical resistivity tomography is not effective in localizing the pipes because they may appear resistive at low frequencies. This is due to the polarization of the metallic pipes which blocks the charge carriers at its external boundaries.

  15. Spontaneous rupture of ovarian cystadenocarcinoma: pre- and post-rupture computed tomography evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Salvadori, Priscila Silveira; Atzingen, Augusto Castelli von; D' Ippolito, Giuseppe [Universidade Federal de Sao Paulo (EPM/UNIFESP), Sao Paulo, SP (Brazil). Escola Paulista de Medicina; Bomfim, Lucas Novais [Universidade Tiradentes (UNIT), Maceio, AL, (Brazil)

    2015-09-15

    Epithelial ovarian tumors are the most common malignant ovarian neoplasms and, in most cases, eventual rupture of such tumors is associated with a surgical procedure. The authors report the case of a 54-year-old woman who presented with spontaneous rupture of ovarian cystadenocarcinoma documented by computed tomography, both before and after the event. In such cases, a post-rupture staging tends to be less favorable, compromising the prognosis. (author)

  16. Global catalog of earthquake rupture velocities shows anticorrelation between stress drop and rupture velocity

    Science.gov (United States)

    Chounet, Agnès; Vallée, Martin; Causse, Mathieu; Courboulex, Françoise

    2018-05-01

    Application of the SCARDEC method provides the apparent source time functions together with seismic moment, depth, and focal mechanism, for most of the recent earthquakes with magnitude larger than 5.6-6. Using this large dataset, we have developed a method to systematically invert for the rupture direction and average rupture velocity Vr, when unilateral rupture propagation dominates. The approach is applied to all the shallow (z earthquakes of the catalog over the 1992-2015 time period. After a careful validation process, rupture properties for a catalog of 96 earthquakes are obtained. The subsequent analysis of this catalog provides several insights about the seismic rupture process. We first report that up-dip ruptures are more abundant than down-dip ruptures for shallow subduction interface earthquakes, which can be understood as a consequence of the material contrast between the slab and the overriding crust. Rupture velocities, which are searched without any a-priori up to the maximal P wave velocity (6000-8000 m/s), are found between 1200 m/s and 4500 m/s. This observation indicates that no earthquakes propagate over long distances with rupture velocity approaching the P wave velocity. Among the 23 ruptures faster than 3100 m/s, we observe both documented supershear ruptures (e.g. the 2001 Kunlun earthquake), and undocumented ruptures that very likely include a supershear phase. We also find that the correlation of Vr with the source duration scaled to the seismic moment (Ts) is very weak. This directly implies that both Ts and Vr are anticorrelated with the stress drop Δσ. This result has implications for the assessment of the peak ground acceleration (PGA) variability. As shown by Causse and Song (2015), an anticorrelation between Δσ and Vr significantly reduces the predicted PGA variability, and brings it closer to the observed variability.

  17. Acute Iliac Artery Rupture: Endovascular Treatment

    International Nuclear Information System (INIS)

    Chatziioannou, A.; Mourikis, D.; Katsimilis, J.; Skiadas, V.; Koutoulidis, V.; Katsenis, K.; Vlahos, L.

    2007-01-01

    The authors present 7 patients who suffered iliac artery rupture over a 2 year period. In 5 patients, the rupture was iatrogenic: 4 cases were secondary to balloon angioplasty for iliac artery stenosis and 1 occurred during coronary angioplasty. In the last 2 patients, the rupture was secondary to iliac artery mycotic aneurysm. Direct placement of a stent-graft was performed in all cases, which was dilated until extravasation was controlled. Placement of the stent-graft was successful in all the cases, without any complications. The techniques used, results, and mid-term follow-up are presented. In conclusion, endovascular placement of a stent-graft is a quick, minimally invasive, efficient, and safe method for emergency treatment of acute iliac artery rupture, with satisfactory short- and mid-term results

  18. Arthrographic diagnosis of ruptured calcaneofibular ligament. I

    International Nuclear Information System (INIS)

    Vuust, M.

    1980-01-01

    A new projection, oblique axial, is recommended for the arthrography of the acute sprained ankle for the correct diagnosis of a ruptured calcaneofibular ligament. Its value is experimentally confirmed. (Auth.)

  19. Right Diaphragm Spontaneous Rupture: A Surgical Approach

    Directory of Open Access Journals (Sweden)

    Duilio Divisi

    2011-01-01

    Full Text Available We present a case of spontaneous rupture of the diaphragm, characterized by nonspecific symptoms. The rapid diagnosis and appropriate surgical approach led to a positive resolution of the pathology.

  20. Traumatic Gallbladder Rupture Treated by Laparoscopic Cholecystectomy

    Science.gov (United States)

    Egawa, Noriyuki; Ueda, Junji; Hiraki, Masatsugu; Ide, Takao; Inoue, Satoshi; Sakamoto, Yuichiro; Noshiro, Hirokazu

    2016-01-01

    Abstract Gallbladder rupture due to blunt abdominal injury is rare. There are few reports of traumatic gallbladder injury, and it is commonly associated with other concomitant visceral injuries. Therefore, it is difficult to diagnose traumatic gallbladder rupture preoperatively when it is caused by blunt abdominal injury. We report a patient who underwent laparoscopic cholecystectomy after an exact preoperative diagnosis of traumatic gallbladder rupture. A 43-year-old man was admitted to our hospital due to blunt abdominal trauma. The day after admission, abdominal pain and ascites increased and a muscular defense sign appeared. Percutaneous drainage of the ascites was performed, and the aspirated fluid was bloody and almost pure bile. He was diagnosed with gallbladder rupture by the cholangiography using the endoscopic retrograde cholangiopancreatography technique. Laparoscopic cholecystectomy was performed safely, and he promptly recovered. If accumulated fluids contain bile, endoscopic cholangiography is useful not only to diagnose gallbladder injury but also to determine the therapeutic strategy. PMID:27462188

  1. Traumatic Gallbladder Rupture Treated by Laparoscopic Cholecystectomy

    Directory of Open Access Journals (Sweden)

    Noriyuki Egawa

    2016-05-01

    Full Text Available Gallbladder rupture due to blunt abdominal injury is rare. There are few reports of traumatic gallbladder injury, and it is commonly associated with other concomitant visceral injuries. Therefore, it is difficult to diagnose traumatic gallbladder rupture preoperatively when it is caused by blunt abdominal injury. We report a patient who underwent laparoscopic cholecystectomy after an exact preoperative diagnosis of traumatic gallbladder rupture. A 43-year-old man was admitted to our hospital due to blunt abdominal trauma. The day after admission, abdominal pain and ascites increased and a muscular defense sign appeared. Percutaneous drainage of the ascites was performed, and the aspirated fluid was bloody and almost pure bile. He was diagnosed with gallbladder rupture by the cholangiography using the endoscopic retrograde cholangiopancreatography technique. Laparoscopic cholecystectomy was performed safely, and he promptly recovered. If accumulated fluids contain bile, endoscopic cholangiography is useful not only to diagnose gallbladder injury but also to determine the therapeutic strategy.

  2. Development of rupture discs for the FBTR

    International Nuclear Information System (INIS)

    Chetal, S.C.; Raju, C.; Anandkumar, V.; Seetharaman, V.; Rajan, K.K.

    1984-01-01

    Rupture discs are required as a safety device for protecting the secondary sodium circuit and its components against high pressure surges due to accidental water steam leaks in sodium heated steam generator and the consequent sodium water reaction. For identical reasons, rupture discs are also required on the vessels used for decontamination of sodium components. Reverse buckling knife blade concept with austenitic stainless steel disc has been developed for the rupture disc assemblies required for Fast Breeder Test Reactor (FBTR). Hydroforming process without any die has been used for disc fabrication. One rupture disc assembly required for steam generator is undergoing sodium endurance test and has accumulated 4,500 hours. The present status of development work as demonstrated by room temperature experimental results as well as the scope for future work are discussed. (author)

  3. Cognitive frames in psychology: demarcations and ruptures.

    Science.gov (United States)

    Yurevich, Andrey V

    2009-06-01

    As there seems to be a recurrent feeling of crisis in psychology, its present state is analyzed in this article. The author believes that in addition to the traditional manifestations that have dogged psychology since it emerged as an independent science some new features of the crisis have emerged. Three fundamental "ruptures" are identified: the "horizontal" rupture between various schools and trends, the "vertical" rupture between natural science and humanitarian psychology, and the "diagonal" rupture between academic research and applied practice of psychology. These manifestations of the crisis of psychology have recently been compounded by the crisis of its rationalistic foundations. This situation is described in terms of the cognitive systems in psychology which include meta-theories, paradigms, sociodigms and metadigms.

  4. Piping inspection activities at the EPRI NDE Center

    International Nuclear Information System (INIS)

    Ammirato, F.V.

    1988-01-01

    Intergranular stress corrosion cracking (IGSCC) in the primary system of boiling water reactors (BWRs) has been a major reliability issue in recent years. BWR pipe cracking was first reported in 1974 with a low percentage of only small-diameter lines affected. However, with increased plant operating time, the number of reported cracking incidents has risen significantly and in 1982 and 1983 included the large-diameter recirculation lines. With the advent of cracking in large-diameter piping, innovative repair remedies were developed, such as weld overlay for repair (WOR). Although these remedies are effective in extending the service life of piping, they also present challenging NDE problems. The EPRI program for improving piping examination has aimed at systematically resolving the difficulties by optimizing techniques and procedures as well as by developing field-qualified automated examination equipment. The EPRI NDE Center's role has been the evaluation and transfer of the technology necessary to address the current piping examination problems of the nuclear utility industry. These activities normally include the following: technology assessment and improvement; validation through demonstrations and field trials; technology transfer reports, workshops, training, and qualification testing; and acquisition of relevant samples. The activities of the NDE Center are discussed

  5. Aortic ruptures in seat belt wearers.

    Science.gov (United States)

    Arajärvi, E; Santavirta, S; Tolonen, J

    1989-09-01

    Several investigations have indicated that rupture of the thoracic aorta is one of the leading causes of immediate death in victims of road traffic accidents. In Finland in 1983, 92% of front-seat passengers were seat belt wearers on highways and 82% in build-up areas. The mechanisms of rupture of the aorta have been intensively investigated, but the relationship between seat belt wearing and injury mechanisms leading to aortic rupture is still largely unknown. This study comprises 4169 fatally injured victims investigated by the Boards of Traffic Accident Investigation of Insurance Companies during the period 1972 to 1985. Chest injuries were recorded as the main cause of death in 1121 (26.9%) victims, 207 (5.0%) of those victims having worn a seat belt. Aortic ruptures were found at autopsy in 98 victims and the exact information of the location of the aortic tears was available in 68. For a control group, we analyzed 72 randomly chosen unbelted victims who had a fatal aortic rupture in similar accidents. The location of the aortic rupture in unbelted victims was more often in the ascending aorta, especially in drivers, whereas in seat belt wearers the distal descending aorta was statistically more often ruptured, especially in right-front passengers (p less than 0.05). The steering wheel predominated statistically as the part of the car estimated to have caused the injury in unbelted victims (37/72), and some interior part of the car was the most common cause of fatal thoracic impacts in seat belt wearers (48/68) (p less than 0.001). The mechanism of rupture of the aorta in the classic site just distal to the subclavian artery seems to be rapid deceleration, although complex body movements are also responsible in side impact collisions. The main mechanism leading to rupture of the ascending aorta seems to be severe blow to the bony thorax. This also often causes associated thoracic injuries, such as heart rupture and sternal fracture. Injuries in the ascending

  6. Traumatic Fundal Rupture of unscarred Uterus in a Primigravida ...

    African Journals Online (AJOL)

    Background: Uterine rupture is an infrequent but life threatening obstetric emergency. Rupture of previously scarred uterus is often encountered especially in multiparous women, but the traumatic rupture of an unscarred primigravid uterus as presented here is a relatively rare event. We report a case of rupture of an ...

  7. Describing Soils: Calibration Tool for Teaching Soil Rupture Resistance

    Science.gov (United States)

    Seybold, C. A.; Harms, D. S.; Grossman, R. B.

    2009-01-01

    Rupture resistance is a measure of the strength of a soil to withstand an applied stress or resist deformation. In soil survey, during routine soil descriptions, rupture resistance is described for each horizon or layer in the soil profile. The lower portion of the rupture resistance classes are assigned based on rupture between thumb and…

  8. Micromechanical properties and collagen composition of ruptured human achilles tendon

    DEFF Research Database (Denmark)

    Hansen, Philip; Kovanen, Vuokko; Hölmich, Per

    2013-01-01

    The Achilles tendon is one of the strongest tendons in the human body, and yet it frequently ruptures, which is a substantial clinical problem. However, the cause of ruptures remains elusive.......The Achilles tendon is one of the strongest tendons in the human body, and yet it frequently ruptures, which is a substantial clinical problem. However, the cause of ruptures remains elusive....

  9. Bladder And Uterine Rupture At Vaginal Birth After Cesarean Section

    African Journals Online (AJOL)

    Rupture of the uterus is still a major obstetric problem in developing countries. The urinary bladder is often affected especially when a lower segment Caesarean scar ruptures. We present a case of bladder rupture associated with uterine rupture in a patient who was attempting vaginal delivery after two previous Caesarean ...

  10. Biomechanical rupture risk assessment of abdominal aortic aneurysms based on a novel probabilistic rupture risk index.

    Science.gov (United States)

    Polzer, Stanislav; Gasser, T Christian

    2015-12-06

    A rupture risk assessment is critical to the clinical treatment of abdominal aortic aneurysm (AAA) patients. The biomechanical AAA rupture risk assessment quantitatively integrates many known AAA rupture risk factors but the variability of risk predictions due to model input uncertainties remains a challenging limitation. This study derives a probabilistic rupture risk index (PRRI). Specifically, the uncertainties in AAA wall thickness and wall strength were considered, and wall stress was predicted with a state-of-the-art deterministic biomechanical model. The discriminative power of PRRI was tested in a diameter-matched cohort of ruptured (n = 7) and intact (n = 7) AAAs and compared to alternative risk assessment methods. Computed PRRI at 1.5 mean arterial pressure was significantly (p = 0.041) higher in ruptured AAAs (20.21(s.d. 14.15%)) than in intact AAAs (3.71(s.d. 5.77)%). PRRI showed a high sensitivity and specificity (discriminative power of 0.837) to discriminate between ruptured and intact AAA cases. The underlying statistical representation of stochastic data of wall thickness, wall strength and peak wall stress had only negligible effects on PRRI computations. Uncertainties in AAA wall stress predictions, the wide range of reported wall strength and the stochastic nature of failure motivate a probabilistic rupture risk assessment. Advanced AAA biomechanical modelling paired with a probabilistic rupture index definition as known from engineering risk assessment seems to be superior to a purely deterministic approach. © 2015 The Author(s).

  11. Blunt cardiac rupture in a toddler

    Directory of Open Access Journals (Sweden)

    Peep Talving

    2016-08-01

    Full Text Available Blunt cardiac rupture is typically a fatal injury with overall mortality exceeding 90%. Most of the patients never reach the hospital alive. In pediatric patients, only 0.03% of cases following blunt trauma admissions have a cardiac injury. This report presents a rare survivor of 16-months old toddler injured in a domestic accident suffering a right atrial rupture repaired through a median sternotomy. To the best of our knowledge this is the youngest case reported in the literature.

  12. Treatment strategy for ruptured abdominal aortic aneurysms.

    Science.gov (United States)

    Davidovic, L

    2014-07-01

    Rupture is the most serious and lethal complication of the abdominal aortic aneurysm. Despite all improvements during the past 50 years, ruptured abdominal aortic aneurysms are still associated with very high mortality. Namely, including patients who die before reaching the hospital, the mortality rate due to abdominal aortic aneurysm rupture is 90%. On the other hand, during the last twenty years, the number of abdominal aortic aneurysms significantly increased. One of the reasons is the fact that in majority of countries the general population is older nowadays. Due to this, the number of degenerative AAA is increasing. This is also the case for patients with abdominal aortic aneurysm rupture. Age must not be the reason of a treatment refusal. Optimal therapeutic option ought to be found. The following article is based on literature analysis including current guidelines but also on my Clinics significant experience. Furthermore, this article show cases options for vascular medicine in undeveloped countries that can not apply endovascular procedures at a sufficient level and to a sufficient extent. At this moment the following is evident. Thirty-day-mortality after repair of ruptured abdominal aortic aneurysms is significantly lower in high-volume hospitals. Due to different reasons all ruptured abdominal aortic aneurysms are not suitable for EVAR. Open repair of ruptured abdominal aortic aneurysm should be performed by experienced open vascular surgeons. This could also be said for the treatment of endovascular complications that require open surgical conversion. There is no ideal procedure for the treatment of AAA. Each has its own advantages and disadvantages, its own limits and complications, as well as indications and contraindications. Future reductions in mortality of ruptured abdominal aortic aneurysms will depend on implementation of population-based screening; on strategies to prevent postoperative organ injury and also on new medical technology

  13. Endovascular treatment of ruptured splenic artery aneurysm

    DEFF Research Database (Denmark)

    Bjerring, Ole Steen

    2008-01-01

    Splenic artery aneurysms (SAA) are traditionally treated surgically, but endovascular techniques are becoming increasingly popular. A 64 year-old male with chest pain and low blood pressure was admitted under suspicion of AMI. A CT scan showed a 56 mm SAA with signs of rupture. The patient...... was treated with endovascular embolisation of the SAA with coils. Blood pressure and haemoglobin levels were stabilized and the patient was discharged. In the case of rupture the treatment of choice seems to be endovascular....

  14. Splenic artery aneurysm rupture in pregnancy

    International Nuclear Information System (INIS)

    Khan, H.R.; Low, S.; Selinger, M.; Nelson, N.

    2004-01-01

    Rupture of a splenic artery aneurysm, commonly associated with pregnancy is a rare and catastrophic event. We report here a case of a patient in her second pregnancy who presented with a short history of left hypochondriac and epigastric pain, followed by collapse at 32 weeks gestation. Sudden fetal distress lead to emergency caesarean delivery when splenic artery aneurysm rupture was diagnosed. With timely involvement of multidisciplinary personnel both mother and baby survived and had an uneventful recovery. (author)

  15. Plantar Fascia Rupture: Ultrasound to Facilitate Recognition.

    Science.gov (United States)

    Servey, Jessica T; Jonas, Christopher

    2018-01-01

    Plantar fascia rupture in the absence of previous diagnosis of plantar fasciitis, corticosteroid injection, or injury is a rare occurrence with only 7 case reports in the literature since 1978. This is a case of spontaneous plantar fascia rupture in a 38-year-old active-duty US military member with current considerations in musculoskeletal ultrasound, other radiologic imaging, treatment, and followup of this diagnosis. © Copyright 2018 by the American Board of Family Medicine.

  16. MRI findings of achilles tendon rupture

    International Nuclear Information System (INIS)

    Zhang Xuezhe

    2009-01-01

    Objective: To evaluate the MRI findings of achilles tendon rupture. Methods: The MRI data of 7 patients with achilles tendon rupture were retrospectively analysed. All 7 patients were male with the age ranging from 34 to 71 years. Routine MR scanning was performed in axial and sagittal planes, including T 1 WI, T 2 WI and a fat suppression MRI (SPIR). Results: Among 7 patients, complete achilles tendon rupture was seen in 6 cases, partial achilles tendon rupture 1 case. The site of tendon disruption were 2.6-11.0 cm( mean 5.4 cm) proximal to the insertion in the calcaneus. The MRI findings of a partial or complete rupture of the achilles tendon included enlarged and thickened achilles tendon (7 cases), wavy lax achilles tendon (2 cases), discontinuity of some or all of its fibers and intratendinous regions of increased signal intensity (7 cases). In the cases of complete tendon rupture, the size of the tendinous gap varied from 3.0-8.0 mm, which was filled with blood and appeared as edema of increase signal intensity on T 2 WI and SPIR. In all 7 patients, MR scanning showed medium signal intensity (7 cases) on T 1 WI, or medium signal intensity (1 cases), medium-high signal intensity (3 cases ), high signal intensity (3 cases) on T 2 WI, and medium-high signal intensity (2 cases), high signal intensity (5 cases) on fat suppression MRI. The preachilles fat pad showed obscure in 6 cases of complete achilles tendon rupture. Conclusion: MRI is an excellent method for revealing achilles tendon rupture and confirming the diagnosis. (authors)

  17. The 1994 Northridge, California, earthquake: Investigation of rupture velocity, risetime, and high-frequency radiation

    Science.gov (United States)

    Hartzell, S.; Liu, P.; Mendoza, C.

    1996-01-01

    A hybrid global search algorithm is used to solve the nonlinear problem of calculating slip amplitude, rake, risetime, and rupture time on a finite fault. Thirty-five strong motion velocity records are inverted by this method over the frequency band from 0.1 to 1.0 Hz for the Northridge earthquake. Four regions of larger-amplitude slip are identified: one near the hypocenter at a depth of 17 km, a second west of the hypocenter at about the same depth, a third updip from the hypocenter at a depth of 10 km, and a fourth updip from the hypocenter and to the northwest. The results further show an initial fast rupture with a velocity of 2.8 to 3.0 km/s followed by a slow termination of the rupture with velocities of 2.0 to 2.5 km/s. The initial energetic rupture phase lasts for 3 s, extending out 10 km from the hypocenter. Slip near the hypocenter has a short risetime of 0.5 s, which increases to 1.5 s for the major slip areas removed from the hypocentral region. The energetic rupture phase is also shown to be the primary source of high-frequency radiation (1-15 Hz) by an inversion of acceleration envelopes. The same global search algorithm is used in the envelope inversion to calculate high-frequency radiation intensity on the fault and rupture time. The rupture timing from the low- and high-frequency inversions is similar, indicating that the high frequencies are produced primarily at the mainshock rupture front. Two major sources of high-frequency radiation are identified within the energetic rupture phase, one at the hypocenter and another deep source to the west of the hypocenter. The source at the hypocenter is associated with the initiation of rupture and the breaking of a high-stress-drop asperity and the second is associated with stopping of the rupture in a westerly direction.

  18. 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)

  19. Endovascular therapeutic strategies in ruptured intracranial aneurysms

    International Nuclear Information System (INIS)

    Machi, Paolo; Lobotesis, Kyriakos; Vendrell, Jean Francoise; Riquelme, Carlos; Eker, Omer; Costalat, Vincent; Bonafe, Alain

    2013-01-01

    The aim of the present study was to evaluate endovascular techniques used currently which were not available at the time of ISAT inclusion period, such as balloon remodelling and flow-divertion, in order to assess whether these new technologies have improved the endovascular approach outcomes. We present a review of articles, published in major journals, with the aim to evaluate the efficacy and the safety of coiling with balloon remodelling for the treatment of ruptured aneurysms in comparison to coiling performed without such coadjutant techniques. Furthermore, we reviewed publications reporting on the treatment of ruptured aneurysms in the acute phase with the one of the most recent technologies available nowadays: the flow diverting stent. Looking at the recent literature the results regarding ruptured aneurysms treated with balloon assisted coiling (BAC) have shown an improvement in terms of anatomical results and morbi-mortality rates. Case series of ruptured middle cerebral artery (MCA) aneurysms treated by EVT report results similar to those obtained by surgical clipping. Several articles recently report encouraging results in treating ruptured dissecting and blister aneurysms with flow diverters. Questions regarding the best treatment available for ruptured aneurysms are yet to be answered. Hence there is a need for a subsequent trial aiming to answer these unresolved issues

  20. 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.

  1. Wall thinning trend analyses for secondary side piping of Korean NPPs

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. A Failure Estimation Method of Steel Pipe Elbows under In-plane Cyclic Loading

    Directory of Open Access Journals (Sweden)

    Bub-Gyu Jeon

    2017-02-01

    Full Text Available The relative displacement of a piping system installed between isolated and nonisolated structures in a severe earthquake might be larger when without a seismic isolation system. As a result of the relative displacement, the seismic risks of some components in the building could increase. The possibility of an increase in seismic risks is especially high in the crossover piping system in the buildings. Previous studies found that an elbow which could be ruptured by low-cycle ratcheting fatigue is one of the weakest elements. Fatigue curves for elbows were suggested based on component tests. However, it is hard to find a quantitative evaluation of the ultimate state of piping elbows. Generally, the energy dissipation of a solid structure can be calculated from the relation between displacement and force. Therefore, in this study, the ultimate state of the pipe elbow, normally considered as failure of the pipe elbow, is defined as leakage under in-plane cyclic loading tests, and a failure estimation method is proposed using a damage index based on energy dissipation.

  3. A failure estimation method of steel pipe elbows under in-plane cyclic loading

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Park, Dong Uk [Seismic Simulation Tester Center, Pusan National University, Yangsan (Korea, Republic of); Kim, Nam Sik [Dept. of Civil and Environmental Engineering, Pusan National University, Busan (Korea, Republic of)

    2017-02-15

    The relative displacement of a piping system installed between isolated and nonisolated structures in a severe earthquake might be larger when without a seismic isolation system. As a result of the relative displacement, the seismic risks of some components in the building could increase. The possibility of an increase in seismic risks is especially high in the crossover piping system in the buildings. Previous studies found that an elbow which could be ruptured by low-cycle ratcheting fatigue is one of the weakest elements. Fatigue curves for elbows were suggested based on component tests. However, it is hard to find a quantitative evaluation of the ultimate state of piping elbows. Generally, the energy dissipation of a solid structure can be calculated from the relation between displacement and force. Therefore, in this study, the ultimate state of the pipe elbow, normally considered as failure of the pipe elbow, is defined as leakage under in-plane cyclic loading tests, and a failure estimation method is proposed using a damage index based on energy dissipation.

  4. A failure estimation method of steel pipe elbows under in-plane cyclic loading

    International Nuclear Information System (INIS)

    Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Park, Dong Uk; Kim, Nam Sik

    2017-01-01

    The relative displacement of a piping system installed between isolated and nonisolated structures in a severe earthquake might be larger when without a seismic isolation system. As a result of the relative displacement, the seismic risks of some components in the building could increase. The possibility of an increase in seismic risks is especially high in the crossover piping system in the buildings. Previous studies found that an elbow which could be ruptured by low-cycle ratcheting fatigue is one of the weakest elements. Fatigue curves for elbows were suggested based on component tests. However, it is hard to find a quantitative evaluation of the ultimate state of piping elbows. Generally, the energy dissipation of a solid structure can be calculated from the relation between displacement and force. Therefore, in this study, the ultimate state of the pipe elbow, normally considered as failure of the pipe elbow, is defined as leakage under in-plane cyclic loading tests, and a failure estimation method is proposed using a damage index based on energy dissipation

  5. A dynamic film model of the pulsating heat pipe

    International Nuclear Information System (INIS)

    Nikolayev, Vadim S.

    2011-01-01

    This article deals with the numerical modeling of the pulsating heat pipe (PHP) and is based on the film evaporation/condensation model recently applied to the single-bubble PHP (Das et al., 2010, 'Thermally Induced Two-Phase Oscillating Flow Inside a Capillary Tube', Int. J. Heat Mass Transfer, 53(19-20), pp. 3905-3913). The described numerical code can treat the PHP of an arbitrary number of bubbles and branches. Several phenomena that occur inside the PHP are taken into account: coalescence of liquid plugs, film junction or rupture, etc. The model reproduces some of the experimentally observed regimes of functioning of the PHP such as chaotic or intermittent oscillations of large amplitudes. Some results on the PHP heat transfer are discussed. (author)

  6. 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)

  7. 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 fuid mass to the structure. This may lead to serious errors.- Inertial effects from the fluid are not correctly evaluated especially in the case of bended or of non-uniform section pipes. Fluid boundary conditions are simply ignored. - In many practical problems fluid compressibility cannot be negelcted, even in the low frequencies domain which corresponds to efficient excitation by turbulent sources of the flow. This paper presents a method to take into account these efects, 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. (Auth.)

  8. Pipe crawler with extendable legs

    International Nuclear Information System (INIS)

    Zollinger, W.T.

    1992-01-01

    A pipe crawler for moving through a pipe in inchworm fashion having front and rear leg assemblies separated by air cylinders to increase and decrease the spacing between assemblies. Each leg of the four legs of an assembly is moved between a wall-engaging, extended position and a retracted position by a separate air cylinder. The air cylinders of the leg assemblies are preferably arranged in pairs of oppositely directed cylinders with no pair lying in the same axial plane as another pair. Therefore, the cylinders can be as long as a leg assembly is wide and the crawler can crawl through sections of pipes where the diameter is twice that of other sections. The crawler carries a valving system, a manifold to distribute air supplied by a single umbilical air hose to the various air cylinders in a sequence controlled electrically by a controller. The crawler also utilizes a rolling mechanism, casters in this case, to reduce friction between the crawler and pipe wall thereby further extending the range of the pipe crawler. 8 figs

  9. Pipe crawler with extendable legs

    Science.gov (United States)

    Zollinger, W.T.

    1992-06-16

    A pipe crawler for moving through a pipe in inchworm fashion having front and rear leg assemblies separated by air cylinders to increase and decrease the spacing between assemblies. Each leg of the four legs of an assembly is moved between a wall-engaging, extended position and a retracted position by a separate air cylinder. The air cylinders of the leg assemblies are preferably arranged in pairs of oppositely directed cylinders with no pair lying in the same axial plane as another pair. Therefore, the cylinders can be as long as a leg assembly is wide and the crawler can crawl through sections of pipes where the diameter is twice that of other sections. The crawler carries a valving system, a manifold to distribute air supplied by a single umbilical air hose to the various air cylinders in a sequence controlled electrically by a controller. The crawler also utilizes a rolling mechanism, casters in this case, to reduce friction between the crawler and pipe wall thereby further extending the range of the pipe crawler. 8 figs.

  10. Microcomputer generated pipe support calculations

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. RESEARCH ON REDUCING PREMATURITY RUPTURE OF MEMBRANE

    Directory of Open Access Journals (Sweden)

    Maria URSACHI (BOLOTA

    2016-12-01

    Full Text Available The membranes surrounding the amniotic cavity are composed from amnion and chorion, tightly adherent layers which are composed of several cell types, including epithelial cells, trophoblasts cells and mesenchyme cells, embedded in a collagenous matrix. They retain amniotic fluid, secret substances into the amniotic fluid, as well as to the uterus and protect the fetus against upward infections from urogenital tract. Normally, the membranes it breaks during labor. Premature rupture of the amniotic sac (PRAS is defined as rupture of membranes before the onset of labor. Premature rupture of the fetal membrane, which occurs before 37 weeks of gestation, usually, refers to preterm premature rupture of membranes. Despite advances in the care period, premature rupture of membranes and premature rupture of membranes preterm continue to be regarded as serious obstetric complications. On the term 8% - 10% of pregnant women have premature rupture of membranes; these women are at increased risk of intrauterine infections, where the interval between membrane rupture and expulsion is rolled-over. Premature rupture of membranes preterm occurs in approximately 1% of all pregnancies and is associated with 30% -40% of preterm births. Thus, it is important to identify the cause of pre-term birth (after less than 37 completed weeks of "gestation" and its complications, including respiratory distress syndrome, neonatal infection and intraventricular hemorrhage. Objectives: the development of the protocol of the clinical trial on patients with impending preterm birth, study clinical and statistical on the socio-demographic characteristics of patients with imminent preterm birth; clinical condition of patients and selection of cases that could benefit from the application of interventional therapy; preclinical investigation (biological and imaging of patients with imminent preterm birth; the modality therapy; clinical investigation of the effectiveness of short

  12. Seismic response analysis of a piping system subjected to multiple support excitations in a base isolated NPP building

    International Nuclear Information System (INIS)

    Surh, Han-Bum; Ryu, Tae-Young; Park, Jin-Sung; Ahn, Eun-Woo; Choi, Chul-Sun; Koo, Ja Choon; Choi, Jae-Boong; Kim, Moon Ki

    2015-01-01

    Highlights: • Piping system in the APR 1400 NPP with a base isolation design is studied. • Seismic response of piping system in base isolated building are investigated. • Stress classification method is examined for piping subjected to seismic loading. • Primary stress of piping is reduced due to base isolation design. • Substantial secondary stress is observed in the main steam piping. - Abstract: In this study, the stress response of the piping system in the advanced power reactor 1400 (APR 1400) with a base isolation design subjected to seismic loading is addressed. The piping system located between the auxiliary building with base isolation and the turbine building with a fixed base is considered since it can be subjected to substantial relative support movement during seismic events. First, the support responses with respect to the base characteristic are investigated to perform seismic analysis for multiple support excitations. Finite element analyses are performed to predict the piping stress response through various analysis methods such as the response spectrum, seismic support movement and time history method. To separately evaluate the inertial effect and support movement effect on the piping stress, the stress is decomposed into a primary and secondary stress using the proposed method. Finally, influences of the base isolation design on the piping system in the APR 1400 are addressed. The primary stress based on the inertial loading is effectively reduced in a base isolation design, whereas a considerable amount of secondary stress is generated in the piping system connecting a base isolated building with a fixed base building. It is also confirmed that both the response spectrum analysis and seismic support movement analysis provide more conservative estimations of the piping stress compared to the time history analysis

  13. The roentgenographic findings of achilles tendon rupture

    Energy Technology Data Exchange (ETDEWEB)

    Seouk, Kang Hyo; Keun, Rho Yong [Shilla General Hospital, Seoul (Korea, Republic of)

    1999-03-01

    To evaluate the diagnostic value of a lateral view of the ankles in Achilles tendon rupture. We performed a retrospective analysis of the roentgenographic findings of 15 patients with surgically proven Achilles tendon rupture. Four groups of 15 patients(normal, ankle sprain, medial lateral malleolar fracture, and calcaneal fracture) were analysed as reference groups. Plain radiographs were reviewed with regard to Kager's triangle, Arner's sign, Toygar's angle, ill defined radiolucent shadow through the Achilles tendon, sharpness of the anterior margin of Achilles tendon, and meniscoid smooth margin of the posterior skin surface of the ankle. Kager's triangle was deformed and disappeared after rupture of the Achilles tendon in nine patients(60%) with operative verification of the rupture, six patients(40%) had a positive Arner's sign, while none had a diminished Toygars angle. In 13 patients(87%) with a ruptured Achilles tendon, the thickness of this was nonuniform compared with the reference group. The anterior margin of the Achilles tendon became serrated and indistinct in 14 patients(93%) in whom this was ruptured. An abnormal ill defined radiolucent shadow through the Achilles tendon was noted in nine patient(60%), and nonparallelism between the anterior margin of the Achilles tendon and posterior skin surface of the ankle was detected in 11 patients(73%). The posterior skin surface of the ankle had a nodular surface margin in 13 patients(87%). A deformed Kager's triangle and Achilles tendon, and an abnormal ill defined radiolucent shadow through the Achilles tendon in a lateral view of the ankles are important findings for the diagnesis of in diagnosing achilles tendon rupture.

  14. The roentgenographic findings of achilles tendon rupture

    International Nuclear Information System (INIS)

    Seouk, Kang Hyo; Keun, Rho Yong

    1999-01-01

    To evaluate the diagnostic value of a lateral view of the ankles in Achilles tendon rupture. We performed a retrospective analysis of the roentgenographic findings of 15 patients with surgically proven Achilles tendon rupture. Four groups of 15 patients(normal, ankle sprain, medial lateral malleolar fracture, and calcaneal fracture) were analysed as reference groups. Plain radiographs were reviewed with regard to Kager's triangle, Arner's sign, Toygar's angle, ill defined radiolucent shadow through the Achilles tendon, sharpness of the anterior margin of Achilles tendon, and meniscoid smooth margin of the posterior skin surface of the ankle. Kager's triangle was deformed and disappeared after rupture of the Achilles tendon in nine patients(60%) with operative verification of the rupture, six patients(40%) had a positive Arner's sign, while none had a diminished Toygars angle. In 13 patients(87%) with a ruptured Achilles tendon, the thickness of this was nonuniform compared with the reference group. The anterior margin of the Achilles tendon became serrated and indistinct in 14 patients(93%) in whom this was ruptured. An abnormal ill defined radiolucent shadow through the Achilles tendon was noted in nine patient(60%), and nonparallelism between the anterior margin of the Achilles tendon and posterior skin surface of the ankle was detected in 11 patients(73%). The posterior skin surface of the ankle had a nodular surface margin in 13 patients(87%). A deformed Kager's triangle and Achilles tendon, and an abnormal ill defined radiolucent shadow through the Achilles tendon in a lateral view of the ankles are important findings for the diagnesis of in diagnosing achilles tendon rupture

  15. Arthroscintigraphy in suspected rotator cuff rupture

    International Nuclear Information System (INIS)

    Gratz, S.; Behr, T.; Becker, W.; Koester, G.; Vosshenrich, R.; Grabbe, E.

    1998-01-01

    Aim: In order to evaluate the diagnostic efficiency of arthroscintigraphy in suspected rotator cuff ruptures this new imaging procedure was performed 20 times in 17 patients with clinical signs of a rotator cuff lesion. The scintigraphic results were compared with sonography (n=20), contrast arthrography (n=20) and arthroscopy (n=10) of the shoulder joint. Methods: After performing a standard bone scintigraphy with intravenous application of 300 MBq 99m-Tc-methylene diphosphonate (MDP) for landmarking of the shoulder region arthroscintigraphy was performed after an intraarticular injection of 99m-Tc microcolloid (ALBU-RES 400 μCi/5 ml). The application was performed either in direct combination with contrast arthrography (n=10) or ultrasound conducted mixed with a local anesthetic (n=10). Findings at arthroscopical surgery (n=10) were used as the gold standard. Results: In case of complete rotator cuff rupture (n=5), arthroscintigraphy and radiographic arthrography were identical in 5/5. In one patient with advanced degenerative alterations of the shoulder joint radiographic arthrography incorrectly showed a complete rupture which was not seen by arthroscintigraphy and endoscopy. In 3 patients with incomplete rupture, 2/3 results were consistant. A difference was seen in one patient with a rotator cuff, that has been already revised in the past and that suffered of capsulitis and calcification. Conclusion: Arthroscinitgraphy is a sensitive technique for detection of rotator cuff ruptures. Because of the lower viscosity of the active compound, small ruptures can be easily detected, offering additional value over radiographic arthrography and ultrasound, especially for evaluation of incomplete cuff ruptures. (orig.) [de

  16. The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kweon, Young Chul

    2002-01-01

    A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR 1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR 1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet

  17. Spontaneous rupture of the esophagus associated with intramural rupture caused by ingestion of weeding medicine (Lasso)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Keon; Park, Heung Il; Kwun, Chung Sik [Chun Nam University College of Medicine, Kwangju (Korea, Republic of)

    1975-06-15

    This is a report of a case of spontaneous rupture of the esophagus associated with intramural rupture caused by ingestion of weeding medicine for the purpose of suicide in a 27 year old Korean male whose chief complaints were dyspnea, epigastric pain, swallowing disturbance, and hoarseness for 3 days prior to admission. A review of literature is submitted.

  18. Flexibility of trunnion piping elbows

    International Nuclear Information System (INIS)

    Lewis, G.D.; Chao, Y.J.

    1987-01-01

    Flexibility factors and stress indices for piping component such as straight pipe, elbows, butt-welding tees, branch connections, and butt-welding reducers are contained in the code, but many of the less common piping components, like the trunnion elbow, do not have flexibility factors or stress indices defined. The purpose of this paper is to identify the in-plane and out-of-plane flexibility factors in accordance with code procedures for welded trunnions attached to the tangent centerlines of long radius elbows. This work utilized the finite element method as applicable to plates and shells for calculating the relative rotations of the trunnion elbow-ends for in-plane and out-of-plane elbow moment loadings. These rotations are used to derive the corresponding in-plane and out-of-plane flexibility factors. (orig./GL)

  19. Pipe Lines – External Corrosion

    Directory of Open Access Journals (Sweden)

    Dan Babor

    2008-01-01

    Full Text Available Two areas of corrosion occur in pipe lines: corrosion from the medium carried inside the pipes; corrosion attack upon the outside of the pipes (underground corrosion. Electrolytic processes are also involved in underground corrosion. Here the moisture content of the soil acts as an electrolyte, and the ions required to conduct the current are supplied by water-soluble salts (chlorides, sulfates, etc. present in the soil. The nature and amount of these soluble materials can vary within a wide range, which is seen from the varying electrical conductivity and pH (varies between 3 and 10. Therefore the characteristics of a soil will be an important factor in under-ground corrosion.

  20. Large Steel Tank Fails and Rockets to Height of 30 meters - Rupture Disc Installed Incorrectly.

    Science.gov (United States)

    Hedlund, Frank H; Selig, Robert S; Kragh, Eva K

    2016-06-01

    At a brewery, the base plate-to-shell weld seam of a 90-m(3) vertical cylindrical steel tank failed catastrophically. The 4 ton tank "took off" like a rocket leaving its contents behind, and landed on a van, crushing it. The top of the tank reached a height of 30 m. The internal overpressure responsible for the failure was an estimated 60 kPa. A rupture disc rated at < 50 kPa provided overpressure protection and thus prevented the tank from being covered by the European Pressure Equipment Directive. This safeguard failed and it was later discovered that the rupture disc had been installed upside down. The organizational root cause of this incident may be a fundamental lack of appreciation of the hazards of large volumes of low-pressure compressed air or gas. A contributing factor may be that the standard piping and instrumentation diagram (P&ID) symbol for a rupture disc may confuse and lead to incorrect installation. Compressed air systems are ubiquitous. The medium is not toxic or flammable. Such systems however, when operated at "slight overpressure" can store a great deal of energy and thus constitute a hazard that ought to be addressed by safety managers.

  1. Report of the U.S. Nuclear Regulatory Commission Piping Review Committee. Summary and evaluation of historical strong-motion earthquake seismic response and damage to aboveground industrial piping

    International Nuclear Information System (INIS)

    1985-04-01

    The primary purpose of this report is to collect in one reference document the observation and experience that has been developed with regard to the seismic behavior of aboveground, building-supported, industrial-type process piping (similar to piping used in nuclear power plants) in strong-motion earthquakes. The report will also contain observations regarding the response of piping in strong-motion experimental tests and appropriate conclusions regarding the behavior of such piping in large earthquakes. Recommendations are included covering the future design of such piping to resist earthquake motion damage based on observed behavior in large earthquakes and simulated shake table testing. Since available detailed data on the behavior of aboveground (building-supported) piping are quite limited, this report will draw heavily on the observations and experiences of experts in the field. In Section 2 of this report, observed earthquake damage to aboveground piping in a number of large-motion earthquakes is summarized. In Section 3, the available experience from strong-motion testing of piping in experimental facilities is summarized. In Section 4 are presented some observations that attempt to explain the observed response of piping to strong-motion excitation from actual earthquakes and shake table testing. Section 5 contains the conclusions based on this study and recommendations regarding the future seismic design of piping based on the observed strong-motion behavior and material developed for the NPC Piping Review Committee. Finally, in Section 6 the references used in this study are presented. It should be understood that the use of the term piping in this report, in general, is limited to piping supported by building structures. It does not include behavior of piping buried in soil media. It is believed that the seismic behavior of buried piping is governed primarily by the deformation of the surrounding soil media and is not dependent on the inertial response

  2. Enhanced seismic criteria for piping

    International Nuclear Information System (INIS)

    Touboul, F. . E-mail francoise.touboul@cea.fr; Blay, N.; Sollogoub, P.; Chapuliot, S.

    2006-01-01

    In situ or laboratory experiments have shown that piping systems exhibit satisfactory seismic behavior. Seismic motion is not severe enough to significantly damage piping systems unless large differential motions of anchorage are imposed. Nevertheless, present design criteria for piping are very severe and require a large number of supports, which creates overly rigid piping systems. CEA, in collaboration with EDF, FRAMATOME and IRSN, has launched a large R and D program on enhanced design methods which will be less severe, but still conservative, and compatible with defect justification during operation. This paper presents the background of the R and D work on this matter, and CEA proposed equations. Our approach is based on the difference between the real behavior (or the best estimated computed one) with the one supposed by codified methods. Codified criteria are applied on an elastically calculated behavior that can be significantly different from the real one: the effect of plasticity may be very meaningful, even with low incursion in the plastic domain. Moreover, and particularly in piping systems, the elastic follow-up effect affects stress distribution for both seismic and thermal loads. For seismic load, we have proposed to modify the elastic moment limitation, based on the interpretation of experimental results on piping systems. The methods have been validated on more industrial cases, and some of the consequences of the changes have been studied: modification of the drawings and of the number of supports, global displacements, forces in the supports, stability of potential defects, etc. The basic aim of the studies undertaken is to make a decision on the stress classification problem, one that is not limited to seismic induced stresses, and to propose simplified methods for its solution

  3. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  4. Splenic rupture following idiopathic rupture of the urinary bladder presenting as acute abdomen

    Directory of Open Access Journals (Sweden)

    Jurisic D

    2007-01-01

    Full Text Available Idiopathic rupture of the urinary bladder is an uncommon condition and represents less than 1% of all bladder rupture cases. In most of the cases the main etiological factor was heavy alcohol ingestion. A combined injury of the spleen and bladder is a very rare condition that is almost often associated with trauma and foreign bodies. In this paper we present the extremely rare clinical course of acute abdomen caused by a combined spontaneous intraperitoneal injury; spontaneous rupture of the urinary bladder and spleen. According to our opinion, spontaneous bladder rupture caused by bladder distension due to alcohol ingestion led to urinary ascites and abdominal distension. Finally, repeated minor abdominal blunt trauma during everyday life, to a moderately distended abdomen caused a spontaneous splenic rupture in the patient with abnormal coagulation studies.

  5. Spontaneous rupture of the urinary bladder in a woman with radiation cystitis. A case report

    Energy Technology Data Exchange (ETDEWEB)

    Kurizaki, Yoshiki; Ishizuka, Osamu [Kofu Municipal Hospital, Yamanashi (Japan)

    1997-07-01

    A 79-year-old woman was admitted to our hospital with gross hematuria and abdominal pain. She had had a uterine cancer 11 years previously and received 56 Gy {sup 60}Co external irradiation combined with 129 Gy {sup 137}Cs internal irradiation. She had a sign of pan-peritonitis. An emergency operation revealed an intraperitoneal rupture of the dome of the urinary bladder 8 cm in length. Because a primary suturing of the bladder wall was unsuccessful, bilateral cutaneous ureterostomy was performed. Histologically, the ruptured bladder wall showed a mucosal erosion and fibrosis of the muscle layer. (author)

  6. Introduction to Loop Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

    This is the presentation file for the short course Introduction to Loop Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. This course will discuss operating principles and performance characteristics of a loop heat pipe. Topics include: 1) pressure profiles in the loop; 2) loop operating temperature; 3) operating temperature control; 4) loop startup; 4) loop shutdown; 5) loop transient behaviors; 6) sizing of loop components and determination of fluid inventory; 7) analytical modeling; 8) examples of flight applications; and 9) recent LHP developments.

  7. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

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

  8. Energy absorbers as pipe supports

    International Nuclear Information System (INIS)

    Khlafallah, M.Z.; Lee, H.M.

    1985-01-01

    With the exception of springs, pipe supports currently in use are designed with the intent of maintaining their rigidity under load. Energy dissipation mechanisms in these pipe supports result in system damping on the order presented by Code Case N-411 of ASME Section III code. Examples of these energy dissipation mechanisms are fluids and gaps in snubbers, gaps in frame supports, and friction in springs and frame supports. If energy absorbing supports designed in accordance with Code Case N-420 are used, higher additional damping will result

  9. Research and design of hanger and support series of nuclear safety class process piping

    International Nuclear Information System (INIS)

    Mao Chengzhang; Shi Jiemin

    1995-12-01

    Hangers and supports of nuclear safety class piping are an important part of primary system piping in a nuclear power plant. They will directly affect the reliability of operation, the period at construction and the investment for a nuclear power plant. It is an absolutely necessary job for Pakistan Chashma Nuclear Power Plant Project to research and design a series of piping supports in accordance with ASME-III NF. It is also an important designing for developing nuclear power plant later in China. After working over two years, a series of piping supports of nuclear safety class which have 57 types and more than 2460 specifications have been designed. This series is perfect, and can satisfy the requirements of piping final designing for nuclear power plant. This series of hangers and supports is mainly used in the process piping of nuclear safety class 1,2,3. They can also be used in other piping of nuclear safety class and piping with aseismic requirement of non-nuclear safety class

  10. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  11. 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

  12. Performance evaluation of buried pipe installation.

    Science.gov (United States)

    2010-05-01

    The purpose of this study is to determine the effects of geometric and mechanical parameters characterizing the soil structure interaction developed in a buried pipe installation located under roads/highways. The drainage pipes or culverts installed ...

  13. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  14. 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

  15. Corrosion of Spiral Rib Aluminized Pipe : [Summary

    Science.gov (United States)

    2012-01-01

    Large diameter, corrugated steel pipes are a common sight in the culverts that run alongside many Florida roads. Spiral-ribbed aluminized pipe (SRAP) has been widely specified by the Florida Department of Transportation (FDOT) for runoff drainage. Th...

  16. Corrosion of Spiral Rib Aluminized Pipe

    Science.gov (United States)

    2012-08-01

    Large diameter, corrugated steel pipes are a common sight in the culverts that run alongside many Florida roads. Spiral-ribbed aluminized pipe (SRAP) has been widely specified by the Florida Department of Transportation (FDOT) for runoff drainage. Th...

  17. Laboratory exercises on oscillation modes of pipes

    Science.gov (United States)

    Haeberli, Willy

    2009-03-01

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

  18. Ultrasonogram of tubal pregnancy: Unruptured or ruptured?

    International Nuclear Information System (INIS)

    Lee, Hee Jung; Kim, Jung Sik; Lee, Sung Moon; Kim, Hong; Woo, Seong Ku; Yeun, Seong Do; Lee, Doo Ryong

    1993-01-01

    To evaluate the usefulness of ultrasonography in the differentiation of the unruptured ones, we performed prospective ultrasonographic study in 56 patients with tubal pregnancy and subsequently correlated with the surgical (45 cases) and laparoscopic (11 cases) findings. The criterion for unruptured tubal pregnancy was the visualization of extra uterine gestational sac or a mass that was separated from the ipsilateral ovary. All cases were confirmed surgically or laparoscopically within 48 hours after sonographic examination. Forty five (80%)cases were proved as unruptured, tubal pregnancy, while 11 (20%) as ruptured one. Extra uterine gestational sac was seen in 31/45 (69%) in unruptured group and 2/11 (18%) in ruptured group. A mass separated from the ipsilateral ovary was seen in 7 unruptured tubal pregnancies and 2 ruptured pregnancies. A mass that was not separated from the ipsilateral ovary was seen in 13 cases. Seven of these were ruptured pregnancies, while 6 were unrupture dones. Using our criteria, the unruptured tubal pregnancy was diagnosed with a sensitivity of 84%, a specificity of 64%, an accuracy of 80%, a positive predictive value of 90% and a negative predictive value of 50%. We conclude that ultrasonography is useful in the differentiation of the unruptured tubal pregnancy from the ruptured one

  19. Diagnosis and Follow-up US Evaluation of Ruptures of the Medial Head of the Gastrocnemius

    International Nuclear Information System (INIS)

    Kwak, Hyo-Sung; Han, Young-Min; Lee, Sang-Yong; Kim, Ki-Nam; Chung, Gyung Ho

    2006-01-01

    The purpose of this study was to demonstrate the ultrasonographic (US) findings of rupture and the healing process of the medial head of the gastrocnemius ('Tennis Leg'). Twenty-two patients (age range: 30 to 45 years) with clinically suspected ruptures of the medial head of the gastrocnemius were referred to us for US examination. All the patients underwent US of the affected limb and the contralateral asymptomatic limb. Follow-up clinical evaluation and US imaging of all patients were performed at two-week intervals during the month after injury and at one-month intervals during the following six months. Of the 22 patients who had an initial US examination after their injury, partial rupture of the medial head of the gastrocnemius muscle was identified in seven patients (31.8%); the remaining 15 patients were diagnosed with complete rupture. Fluid collection between the medial head of the gastrocnemius and the soleus muscle was identified in 20 patients (90.9%). The thickness of the fluid collection, including the hematoma in the patients with complete rupture (mean: 9.7 mm), was significantly greater than that seen in the patients with partial tear (mean: 6.8 mm) (p < 0.01). The primary union of the medial head of the gastrocnemius with the soleus muscle in all the patients with muscle rupture and fluid collection was recognized via the hypoechoic tissue after four weeks. Ultrasonography is a useful imaging modality for the diagnosis and follow-up examination for the patients suffering with rupture of the medial head of the gastrocnemius

  20. Analysis of Municipal Pipe Network Franchise Institution

    Science.gov (United States)

    Yong, Sun; Haichuan, Tian; Feng, Xu; Huixia, Zhou

    Franchise institution of municipal pipe network has some particularity due to the characteristic of itself. According to the exposition of Chinese municipal pipe network industry franchise institution, the article investigates the necessity of implementing municipal pipe network franchise institution in China, the role of government in the process and so on. And this offers support for the successful implementation of municipal pipe network franchise institution in China.