Robots in pipe and vessel inspection: past, present, and future

International Nuclear Information System (INIS)

Mueller, T.A.; Tyndall, J.F.

1984-01-01

Over the past several decades, remotely operated scanners have been employed to inspect piping and pressure vessels. These devices in their early forms were manually controlled manipulators functioning as mere extensions of the operator. With the addition of limit sensing, speed control, and positional feedback and display, the early manipulators became primitive robots. By adding computer controls with their degree of intelligence to the devices, they achieved the status of robots. Future applications of vision, adaptive control, proximity sensing, and pattern recognition will bring these devices to a level of intelligence that will make automated robotic inspection of pipes and pressure vessels a true reality

Leak detector for reactor pressure vessel

International Nuclear Information System (INIS)

Morimoto, Mikio.

1991-01-01

A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

The Assembly and Test of Pressure Vessel for Irradiation

International Nuclear Information System (INIS)

Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

2009-01-01

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

The Assembly and Test of Pressure Vessel for Irradiation

Energy Technology Data Exchange (ETDEWEB)

Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kennedy, Timothy C. [Oregon State University, Corvallis (United States)

2009-02-15

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

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

International Nuclear Information System (INIS)

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

1985-10-01

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

Structural features and in-service inspection of the LTHR-200 pressure vessel

International Nuclear Information System (INIS)

Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

1993-01-01

LTHR-200 is a low temperature district-heating reactor. It adopts double-shell design pressure vessel and metal containment. Because of the safety and structural features of the reactor, the in-service inspection of the pressure vessel can be simplified greatly. LTHR-200 is an integrated arrangement. Both its core components and the main heat exchangers are contained in the reactor pressure vessel. The coolant of the main loop is run by a full-power natural circulation and there need no main pumps and pipes. Thus, the reactor pressure vessel constitutes the pressure boundary of the reactor's main loop coolant. In regard to these features, a small-sized containment is designed for the reactor. The metal safety container with a small volume is placed closely around the reactor pressure vessel. Outside the metal containment, there is a large reinforced concrete construction for the reactor. Their main operation and design parameters are as follows: The pressure vessel: operation pressure = 2.4 MPa; design pressure = 3.0 MPa; design temperature = 250 deg C; 40 year fast neutron (E>1MeV) fluence in the belt-line region = < 10E16n/cm; internal diameter = 5000 mm; material SA516-70; shell thickness 65 mm; The metal containment: maximum operation pressure = 1.8 MPa; design pressure = 1.8 MPa; design temperature = 250 deg. C; upper internal diameter 7000 mm; lower internal diameter = 5600 mm; material = SA516-70; shell thickness, upper part = 80 mm; lower part = 50 mm. All penetrating pipes through the pressure vessel are located at the top penetration section of the shell. All the internal diameters of penetrating pipes are less than 50 mm. Inside and outside the metal containment wall respectively, isolating valves are connected to the reactor coolant pipe which passes through the containment. These two isolating valves use different driving methods. Every penetrating part of the reactor construction uses a proper form of structure according to safety requirements

Elastic-plastic analysis of part-through crack propagation in piping and pressure vessels

International Nuclear Information System (INIS)

Souza, L.A. de; Ebecken, N.F.F.

1986-01-01

The shell structures, often used in the construction of reservoirs, pipings, pressure vessels, nuclear power plants, etc, with part-through crack along its thickness, are analysed, using a computer system developed by the finite element method. The surface is discretized with three-dimensional quadratic elements, degenerated in its mid-surface, such the fracture is simulated by scalar elements (non linear springs). The results are analysed by the stress intensity factor K Sub(I) and the strain energy release rate, which is known as J-integral. The analysis is performed in the elastic and elastic-plastic regime. The basic hipothesis and the formulation adopted in the derivation of the scalar elements are also shown. (Author) [pt

ITER cryostat main chamber and vacuum vessel pressure suppression system design

International Nuclear Information System (INIS)

Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

1999-03-01

Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

ITER cryostat main chamber and vacuum vessel pressure suppression system design

Energy Technology Data Exchange (ETDEWEB)

Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

1999-03-01

Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

Experimental study on pressure wave propagation through the open end of pipe

International Nuclear Information System (INIS)

Yoshida, K.; Kumagai, H.

1994-01-01

The steam generators of a double pool type liquid metal fast breeder reactor (LMFBR) are used in a large sodium pool which is formed between the primary vessel and the secondary vessel and accommodates the entire secondary heat transport system. Therefore, if there is a sodium-water reaction event in the steam generator, it becomes important to evaluate the pressure rises at the walls of the primary and secondary vessels as well as those at the other secondary components. An experimental study was performed, focusing on the propagation of the initial pressure spike of the-sodium-water reaction from the bottom end of the steam generator to the sodium pool. Pressure wave propagation from inside of a pipe to an open space through the pipe end was measured. Two kinds of pressure propagation media, water and air, ensured a wide range of experimental conditions. The experimental results revealed that the pressure attenuation at the open end of a pipe can be put in order using the concept of inertial length, and that the dimensionless inertial length, i.e., the inertial length divided by the half wave length of the pressure pulse, is proportional to the square of the dimensionless diameter. These results provide a prediction method for a pressure rise by the initial pressure spike in the secondary sodium pool of the Double Pool LMFBR

Dictionary of pressure vessel and piping technology. English-German. Woerterbuch der Druckbehaelter- und Rohrleitungstechnik. Englisch-Deutsch

Energy Technology Data Exchange (ETDEWEB)

Jentgen, L; Schmitz, H P

1986-01-01

A specialised dictionary has been compiled containing the appropriate English and German terms in the following technical fields: materials science, welding, destructive and non-destructive testing, thermal and mass transfer, the design and construction in particular of pressure vessels, tanks, heat exchangers, piping, expansion joints, valves, and components associated with the above fields. This dictionary is the result of many years spent in evaluating technical terminology from the relevant American and British regulations, technical rules, standards, and specifications (see bibliography) and correlating these with the terminology of comparable German regulations, rules and standards, together with the essential technical literature.

Assessment of integrity for the pressure vessel internals of PWRs under blowdown loadings

International Nuclear Information System (INIS)

Geiss, M.; Benner, J.; Ludwig, A.

1984-01-01

In safety analysis of pressurized water reactors the loss-of-coolant accident plays a central role. Thereby a sudden break of a cold primary coolant pipe close to the reactor pressure vessel is postulated. The sudden pressure release of the primary system (blowdown) causes high dynamic loading on the pressure vessel internals. The resulting deformations must not impair shut down of the reactor and decay heat removal in an inadmissible way. For this assessment a blowdown analysis for a 1300 MW pressurized water reactor is carried out. These investigations are completed with a detailed stress analysis for the highly loaded core barrel clamping. The results show that the reactor pressure vessel internals are able to withstand blowdown loading. Even in case of a sudden and complete break of the primary coolant pipe the loading has to be twice as high to endanger the structural integrity. (orig.) [de

Leak detection device for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Ikeda, Jun.

1988-01-01

Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

The criteria of fracture in the case of the leak of pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Habil; Ziliukas, A.

1997-04-01

In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

Pressure thermal shock analysis for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

2015-01-01

The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

Dictionary of pressure vessel and piping technology. German-English. Woerterbuch der Druckbehaelter- und Rohrleitungstechnik. Deutsch-Englisch

Energy Technology Data Exchange (ETDEWEB)

Schmitz, H P

1987-01-01

This dictionary is the result of many years of evaluation of technical terminology taken from the salient non-German rules, regulations, standards and specifications such as ANSI, API, ASME, ASNT, ASTM, BSI, EJMA, TEMA, and WRC (see bibliography) and of comparing these with the corresponding German rules, regulations, etc., as well as examining relevant technical documentation. This dictionary fills the gap left by existing dictionaries. The following specialized factors are given special attention: pressure vessels, tanks, heat exchangers, piping, valves and fittings, expansion joints, flanges, giving particular consideration to the fields of materials, welding, strength calculation, design and construction, fracture mechanics, destructive and non-destructive testing, as well as heat and mass transfer.

Towards a new pressure vessel standard in the European Union

International Nuclear Information System (INIS)

Osweiller, F.

1995-01-01

Since 1990 the European Commission has been preparing a new Directive which will regulate the Pressure Equipment sector in the countries of the European Union. CEN Standards devoted to pressure vessels, piping, boilers, are currently being drawn up to complete and implement this Directive. This paper focuses on the European Unfired Pressure Vessel Standard (EPVS) which is in course of development under the responsibility of CEN/TC54. The main aspects of the Standard are outlined: general structure, materials, design, fabrication, inspection and testing. The link with the European Directive is explained in connection with regulatory aspects: conformity assessment, essential safety requirements, classes of vessels, notified bodies, EC mark, status of the standard

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

A ductile fracture mechanics methodology for predicting pressure vessel and piping failure

International Nuclear Information System (INIS)

Landes, J.D.; Zhou, Z.

1991-01-01

This paper reports on a ductile fracture methodology based on one used more generally for the prediction of fracture behavior that was applied to the prediction of fracture behavior in pressure vessel and piping components. The model uses the load versus displacement record from a fracture toughness test to develop inputs for predicting the behavior of the structural component. The principle of load separation is used to convert the test record into two pieces of information, calibration functions which describe the structural deformation behavior and fracture toughness which describes the response of a crack-like flaw to the loading. These calibration functions and fracture toughness values which relate to the test specimen are then transformed to those appropriate to the structure. Often in this step computation procedures could be used but are not always necessary. The calibration functions and fracture for the structure are recombined to predict a load versus displacement behavior for the structure. The input for the model was generated from tests of compact specimen geometries; this geometry is often used for fracture toughness testing. The predictions were done for five model structures

Factors in the fail safe approach to pressure vessel assessment

International Nuclear Information System (INIS)

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

1978-01-01

The 'leak before break' concept is described in the context of pressure vessel assessment. The factors which determine whether a pipe containing an axial flaw will leak or break under pressure loading are discussed using a post-yield fracture mechanics method. A model is used in which it is assumed that initially the ligament beneath the flaw fails to form a full thickness defect in the pipe. The stability of the full thickness defect at the pressure causing ligament failure is then examined to ascertain whether it would remain at the snap through length or would propagate by fast fracture, to form a leak or a break. The method is used to analyse the results of a series of pipe rupture tests, and it is found that a distinction between leaks and breaks is achieved. (author)

Problems in manufacturing and transport of pressure vessels of integral reactors

International Nuclear Information System (INIS)

Kralovec, J.

1997-01-01

Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, i.e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters, heights and thick walls and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabilities. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment, machining, shop-handling, non-destructive testing, hydraulic pressure tests etc. Tile transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport each of them having its advantages and limitations. Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper. (author)

Strength-toughness requirements for thick walled high pressure vessels

International Nuclear Information System (INIS)

Kapp, J.A.

1990-01-01

The strength and toughness requirements of materials for use in high pressure vessels has been the subject of some discussion in the meetings of the Materials Task Group of the Special Working Group High Pressure Vessels. A fracture mechanics analysis has been performed to theoretically establish the required toughness for a high pressure vessel. This paper reports that the analysis performed is based on the validity requirement for plane strain fracture of fracture toughness test specimens. This is that at the fracture event, the crack length, uncracked ligament, and vessel length must each be greater than fifty times the crack tip plastic zone size for brittle fracture to occur. For high pressure piping applications, the limiting physical dimension is the uncracked ligament, as it can be assumed that the other dimensions are always greater than fifty times the crack tip plastic zone. To perform the fracture mechanics analysis several parameters must be known: these include vessel dimensions, material strength, degree of autofrettage, and design pressure. Results of the analysis show, remarkably, that the effects of radius ratio, pressure and degree of autofrettage can be ignored when establishing strength and toughness requirements for code purposes. The only parameters that enter into the calculation are yield strength, toughness and vessel thickness. The final results can easily be represented as a graph of yield strength against toughness on which several curves, one for each vessel thickness, are plotted

Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

2008-09-15

This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

Ratcheting study in pressurized piping components under cyclic loading at room temperature

International Nuclear Information System (INIS)

Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.

2006-07-01

The nuclear power plant piping components and systems are often subjected to reversing cyclic loading conditions due to various process transients, seismic and other events. Earlier the design of piping subjected to seismic excitation was based on the principle of plastic collapse. It is believed that during such events, fatigue-ratcheting is likely mode of failure of piping components. The 1995 ASME Boiler and Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. Experimental and analytical studies are carried out to understand this failure mechanism. The biaxial ratcheting characteristics of SA 333, Gr. 6 steel and SS 304 stainless steel at room temperature are investigated in the present work. Experiments are carried out on straight pipes subjected to internal pressure and cyclic bending load applied in a three point and four point bend test configurations. A shake table test is also carried out on a pressurized elbow by applying sinusoidal base excitation. Analytical simulation of ratcheting in the piping elements is carried out. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. (author)

Evaluation of burst pressure prediction models for line pipes

Energy Technology Data Exchange (ETDEWEB)

Zhu, Xian-Kui, E-mail: zhux@battelle.org [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States); Leis, Brian N. [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States)

2012-01-15

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487-492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: Black-Right-Pointing-Pointer This paper evaluates different burst pressure prediction models for line pipes. Black-Right-Pointing-Pointer The existing models are categorized into two major groups of Tresca and von Mises solutions. Black-Right-Pointing-Pointer Prediction quality of each model is assessed statistically using a large full-scale burst test database. Black-Right-Pointing-Pointer The Zhu-Leis solution is identified as the best predictive model.

Evaluation of burst pressure prediction models for line pipes

International Nuclear Information System (INIS)

Zhu, Xian-Kui; Leis, Brian N.

2012-01-01

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487–492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: ► This paper evaluates different burst pressure prediction models for line pipes. ► The existing models are categorized into two major groups of Tresca and von Mises solutions. ► Prediction quality of each model is assessed statistically using a large full-scale burst test database. ► The Zhu-Leis solution is identified as the best predictive model.

Response margins investigation of piping dynamic analyses using the independent support motion method and PVRC [Pressure Vessel Research Committee] damping

International Nuclear Information System (INIS)

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

1988-03-01

An evaluation of Independent Support Motion (ISM) response spectrum methods of analysis coupled with the Pressure Vessel Research Committee (PVRC) recommendation for damping, to compute the dynamic component of the seismic response of piping systems, was completed. Response estimates for five piping/structural systems were developed using fourteen variants of the ISM response spectrum method, the Uniform Support Motions response spectrum method and the ISM time history analysis method, all based on the PVRC recommendations for damping. The ISM/PVRC calculational procedures were found to exhibit orderly characteristics with levels of conservatism comparable to those obtained with the ISM/uniform damping procedures. Using the ISM/PVRC response spectrum method with absolute combination between group contributions provided consistently conservative results while using the ISM/PVRC response spectrum method with square root sum of squares combination between group contributions provided estimates of response which were deemed to be acceptable

Multiple shell pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.

1988-01-01

A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

Prevention of catastrophic failure in pressure vessels and pipings

International Nuclear Information System (INIS)

Rintamaa, R.; Wallin, K.; Ikonen, K.; Toerroenen, K.; Talja, H.; Keinaenen, H.; Saarenheimo, A.; Nilsson, F.; Sarkimo, M.; Waestberg, S.; Debel, C.

1989-01-01

The fracture resistance and integrity of pressure-loaded components have been assessed in a Nordic research programme. Experiments were performed to validate the computational fracture assessment analysis. Two tests were also conducted on a large decommissioned pressure vessel from an oil refinery plant. Different fracture assessment methods were developed and subsequently applied to the tested components. Interlaboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finit element calculations and fracture mechanics testing. The transferability of material parameters derived from small specimens with simple crack geometries to more realistic crack geometries in real components has been verified. (author)

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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.

Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

International Nuclear Information System (INIS)

Kim, Kyung Mo; Bang, In Cheol

2017-01-01

Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

Failure pressure of straight pipe with wall thinning under internal pressure

International Nuclear Information System (INIS)

Kamaya, Masayuki; Suzuki, Tomohisa; Meshii, Toshiyuki

2008-01-01

The failure pressure of pipe with wall thinning was investigated by using three-dimensional elastic-plastic finite element analyses (FEA). With careful modeling of the pipe and flaw geometry in addition to a proper stress-strain relation of the material, FEA could estimate the precise burst pressure obtained by the tests. FEA was conducted by assuming three kinds of materials: line pipe steel, carbon steel, and stainless steel. The failure pressure obtained using line pipe steel was the lowest under the same flaw size condition, when the failure pressure was normalized by the value of unflawed pipe defined using the flow stress. On the other hand, when the failure pressure was normalized by the results of FEA obtained for unflawed pipe under various flaw and pipe configurations, the failure pressures of carbon steel and line pipe steel were almost the same and lower than that of stainless steel. This suggests that the existing assessment criteria developed for line pipe steel can be applied to make a conservative assessment of carbon steel and stainless steel

Containment vessel for a nuclear reactor

International Nuclear Information System (INIS)

Yamanari, Sh.; Horiuchi, T.; Sugisaki, T.; Tominaga, K.

1985-01-01

A containment vessel for a nuclear reactor having a dry well for mounting therein a pressure vessel for containing the nuclear reactor, a pressure suppressing chamber having a pool of coolant therein, and a vent pipe device for releasing therethrough into the pool of coolant within the pressure suppressing chamber steam which will be produced as a result of the occurrence of an accident and escape into the dry well. The vent pipe device includes a plurality of vent pipe members inserted in the pool of coolant within the pressure suppressing chamber and each having at least one exhaust port opening in the coolant. The vent pipe members are divided into a plurality of groups in such a manner that the vent pipe members of different groups differ from one another in the length of submerged portions of the vent pipe members interposed between the liquid of the coolant within the pressure suppressing chamber and the exhaust ports of the vent pipe members

Fabrication of High Temperature and High Pressure Vessel for the Fuel Test

International Nuclear Information System (INIS)

Park, Kook Nam; Lee, Jong Min; Sim, Bong Shick; Shon, Jae Min; Ahn, Seung Ho; Yoo, Seong Yeon

2007-01-01

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR and CANDU nuclear power plants has been developed and installed in HANARO, KAERI. It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is located inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

Local and global collapse pressure of longitudinally flawed pipes and cylindrical vessels

International Nuclear Information System (INIS)

Staat, M.

2005-01-01

Limit loads can be calculated with the finite element method (FEM) for any component, defect geometry, and loading. FEM suggests that published long crack limit formulae for axial defects under-estimate the burst pressure for internal surface defects in thick pipes while limit loads are not conservative for deep cracks and for pressure loaded crack-faces. Very deep cracks have a residual strength, which is modelled by a global collapse load. These observations are combined to derive new analytical local and global collapse loads. The global collapse loads are close to FEM limit analyses for all crack dimensions

Stress index development for piping with trunnion attachment under pressure and moment loading

International Nuclear Information System (INIS)

Lee, D. H.; Kim, J. M.; Park, S. H.

1997-01-01

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per section III of the ASME boiler and pressure vessel code from which the Primary (B 1 ), Secondary (C 1 ) and Peak (K 1 ) stress indices for pressure, the Primary (B 2 ), Secondary (C 2 ) and Peak (K 2 ) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage. (author)

Special enclosure for a pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.; Wedellsborg, U.W.

1993-01-01

A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom

Pressure vessel design manual

CERN Document Server

Moss, Dennis R

2013-01-01

Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

Consequence evaluation of hypothetical reactor pressure vessel support failure

International Nuclear Information System (INIS)

Lu, S.C.; Holman, G.S.; Lambert, H.E.

1991-01-01

This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

International pressure vessels and piping codes and standards. Volume 2: Current perspectives; PVP-Volume 313-2

International Nuclear Information System (INIS)

Rao, K.R.; Asada, Yasuhide; Adams, T.M.

1995-01-01

The topics in this volume include: (1) Recent or imminent changes to Section 3 design sections; (2) Select perspectives of ASME Codes -- Section 3; (3) Select perspectives of Boiler and Pressure Vessel Codes -- an international outlook; (4) Select perspectives of Boiler and Pressure Vessel Codes -- ASME Code Sections 3, 8 and 11; (5) Codes and Standards Perspectives for Analysis; (6) Selected design perspectives on flow-accelerated corrosion and pressure vessel design and qualification; (7) Select Codes and Standards perspectives for design and operability; (8) Codes and Standards perspectives for operability; (9) What's new in the ASME Boiler and Pressure Vessel Code?; (10) A look at ongoing activities of ASME Sections 2 and 3; (11) A look at current activities of ASME Section 11; (12) A look at current activities of ASME Codes and Standards; (13) Simplified design methodology and design allowable stresses -- 1 and 2; (14) Introduction to Power Boilers, Section 1 of the ASME Code -- Part 1 and 2. Separate abstracts were prepared for most of the individual papers

Radioactive liquid containing vessel

International Nuclear Information System (INIS)

Sakurada, Tetsuo; Kawamura, Hironobu.

1993-01-01

Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

Limit loads for piping branch junctions under internal pressure and in-plane bending-Extended solutions

International Nuclear Information System (INIS)

Kim, Yun-Jae; Lee, Kuk-Hee; Park, Chi-Yong

2008-01-01

The authors have previously proposed plastic limit load solutions for thin-walled branch junctions under internal pressure and in-plane bending, based on finite element (FE) limit loads resulting from three-dimensional (3-D) FE limit analyses using elastic-perfectly plastic materials [Kim YJ, Lee KH, Park CY. Limit loads for thin-walled piping branch junctions under internal pressure and in-plane bending. Int J Press Vessels Piping 2006;83:645-53]. The solutions are valid for ratios of the branch-to-run pipe radius and thickness from 0.4 to 1.0, and for the mean radius-to-thickness ratio of the run pipe from 10.0 to 20.0. Moreover, the solutions considered the case of in-plane bending only on the branch pipe. This paper extends the previous solutions in two aspects. Firstly, plastic limit load solutions are given also for in-plane bending on the run pipe. Secondly, the validity of the proposed solutions is extended to ratios of the branch-to-run pipe radius and thickness from 0.0 to 1.0, and the mean radius-to-thickness ratio of the run pipe from 5.0 to 20.0. Comparisons with FE results show good agreement

Sodium steam generator within which are inlet and outlet ducts with pipe bundles in vessel

International Nuclear Information System (INIS)

1980-01-01

The sodium steam generator with internal flow ducts for inlet and outlet to a vessel are provided as pipe bundles in the form of helically wound concentric layers terminating in inlet and outlet connections with chambers, characterised in that within the vessel, the pipe pieces which are connected to the pipe windings with the said vessel are arranged in substantially radially aligned rows so that each row measured in the circumferential direction at least on one side is at a spacing from the following row sufficiently large that between the rows or groups of rows an open sector is provided. (G.C.)

Metallurgical failure investigation of a pipe connector fracture of an expansion vessel

International Nuclear Information System (INIS)

Neidel, Andreas

2016-01-01

A pipe connector of an expansion vessel of a safety heat exchanger was torn off in a test facility's natural gas compressor. From a material point of view, the cause of the damage is a fatigue fracture induced by pulsating bending stress. The fatigue fracture originated from both, the pipe's outer surface as well as from its inner surface, which is consistent with the given stress situation (pulsating bending stress). Material defects or welding-induced flaws were not observed. Corrosion, wear, or thermal overload which may have promoted the damage, were not observed either. The primary cause was a major design error. Cases of dynamic load were obviously not duly taken into account during designing, so that the free-swinging mass of the expansion vessel which was mounted to a pipe of a diameter of only half an inch and, furthermore, installed in an angle of 45 (additional static preload.), could cause the fatigue failure induced by pulsating bending stress in the zone of highest stresses at the transition of the expansion vessel and the the pipe connector due to dynamic operating loads which always occur in plants like these.

Apparatus and method for determining stress and strain in pipes, pressure vessels, structural members and other deformable bodies

International Nuclear Information System (INIS)

Vachon, R.I.; Ranson, W.F.

1987-01-01

A method and apparatus for measuring stress and strain associated with a pipe, pressurized vessel, structural member or deformable body containing a flaw or stress concentration utilizes a laser beam to illuminate a surface being analyzed and an optical data digitizer to sense a signal provided by a speckle pattern produced by the light beam reflected from the illuminated surface. One signal is received from the surface in a reference condition and subsequent signals are received from the surface after surface deformation. The optical data digitizer provides the received signal to an image processor, and the processor stores the signals and correlates the deformed image received with the reference image and then sends this correlated information to a minicomputer which performs mathematical analyses of the signal to determine stress and strain associated with the surface. The apparatus is constructed as one integral unit, and further includes a digital and tape display, as well as a television monitor and an electro-optic range indicator. (author) 15 figs

Pressure vessel design

International Nuclear Information System (INIS)

Annaratone, D.

2007-01-01

This book guides through general and fundamental problems of pressure vessel design. It moreover considers problems which seem to be of lower importance but which turn out to be crucial in the design phase. The basic approach is rigorously scientific with a complete theoretical development of the topics treated, but the analysis is always pushed so far as to offer concrete and precise calculation criteria that can be immediately applied to actual designs. This is accomplished through appropriate algorithms that lead to final equations or to characteristic parameters defined through mathematical equations. The first chapter describes how to achieve verification criteria, the second analyzes a few general problems, such as stresses of the membrane in revolution solids and edge effects. The third chapter deals with cylinders under pressure from the inside, while the fourth focuses on cylinders under pressure from the outside. The fifth chapter covers spheres, and the sixth is about all types of heads. Chapter seven discusses different components of particular shape as well as pipes, with special attention to flanges. The eighth chapter discusses the influence of holes, while the ninth is devoted to the influence of supports. Finally, chapter ten illustrates the fundamental criteria regarding fatigue analysis. Besides the unique approach to the entire work, original contributions can be found in most chapters, thanks to the author's numerous publications on the topic and to studies performed ad hoc for this book. (orig.)

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

International Nuclear Information System (INIS)

Xu Bin; Feng Quanke; Yu Xiaoling

2008-01-01

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

Stress analysis of LOFT containment vessel attachments for the mainsteam and feedwater piping support structures

International Nuclear Information System (INIS)

Finicle, D.P.

1977-01-01

The LOFT Containment Vessel attachments for the Mainsteam and Feedwater Piping Support Structures have been analyzed for operating and faulted loading conditions. This report contains the analysis of the connections to the containment vessel for the most current design and loading. Also contained in this report is the analysis of the piping supports

Inspection technology for high pressure pipes

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae H.; Lee, Jae C.; Eum, Heung S.; Choi, Yu R.; Moon, Soon S.; Jang, Jong H

2000-02-01

Various kinds of defects are likely to be occurred in the welds of high pressure pipes in nuclear power plants. Considering the recent accident of Zuruga nuclear power plant in Japan, reasonable policy is strongly requested for the high pressure pipe integrity. In this study, we developed the technologies to inspect pipe welds automatically. After development of scanning robot prototype in the first research year, we developed and implemented the algorithm of automatic tracking of the scanning robot along the weld line of the pipes. We use laser slit beam on weld area and capture the image using digital camera. Through processing of the captures image, we finally determine the weld line automatically. In addition, we investigated a new technology on micro systems for developing micro scanning robotic inspection of the pipe welds. The technology developed in this study is being transferred to the industry. (author)

Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

International Nuclear Information System (INIS)

Wang, G.Y.; Shin, Y.W.; Moody, F.J.

1991-01-01

This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

46 CFR 32.50-25 - Cargo pumps and piping on tank vessels constructed prior to November 10, 1936-TB/ALL.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Cargo pumps and piping on tank vessels constructed prior... SECURITY TANK VESSELS SPECIAL EQUIPMENT, MACHINERY, AND HULL REQUIREMENTS Pumps, Piping, and Hose for Cargo Handling § 32.50-25 Cargo pumps and piping on tank vessels constructed prior to November 10, 1936—TB/ALL...

Defect Depth Measurement of Straight Pipe Specimen Using Shearography

International Nuclear Information System (INIS)

Chang, Ho Seob; Kim, Kyung Suk

2012-01-01

In the nuclear industry, wall thinning defect of straight pipe occur the enormous loss in life evaluation and safety evaluation. To use non-destructive technique, we measure deformation, vibration, defect evaluation. But, this techniques are a weak that is the measurement of the wide area is difficult and the time is caught long. In the secondary side of nuclear power plants mostly used steel pipe, artificiality wall thinning defect make in the side and different thickness make to the each other, wall thinning defect part of deformation measure by using shearography. In addition, optical measurement through deformation, vibration, defect evaluation evaluate pipe and thickness defects of pressure vessel is to evaluate quantitatively. By shearography interferometry to measure the pipe's internal wall thinning defect and the variation of pressure use the proposed technique, the quantitative defect is to evaluate the thickness of the surplus. The amount of deformation use thickness of surplus prediction of the actual thickness defect and approximately 7 percent error by ensure reliability. According to pressure the amount of deformation and the thickness of the surplus through DB construction, nuclear power plant pipe use wall thinning part soundness evaluation. In this study, pressure vessel of thickness defect measure proposed nuclear pipe of wall thinning defect prediction and integrity assessment technology development. As a basic research defected theory and experiment, pressure vessel of advanced stability and soundness and maintainability is expected to contribute foundation establishment

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

Hydrogen storage in insulated pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

1998-08-01

Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

Design of pressure vessels. Part 2

International Nuclear Information System (INIS)

Grandemange, J.M.

2008-01-01

This document deals with the classification of stresses, necessary for the implementation of the mechanical code criteria defined for the pressure vessels of PWR-type reactors. It describes the general approach of design, analysis, and in-service monitoring, the regulatory tests and the modalities of equivalence between industrial construction codes. Content: 1 - damage modes and stresses classification: context, general approach, example of application; 2 - from the design stage to the in-service monitoring: liabilities, design conditions, materials choice and dimensioning, analysis, particular case of pipes and valve parts, in-service monitoring; 3 - regulatory tests: context, tests prescribed by the design and construction rules of PWR mechanical components (RCC-M); 4 - equivalence possibilities between codes: codes for nuclear reactor equipments, convergence between industrial codes and standards; 5 - conclusion. (J.S.)

46 CFR 52.01-105 - Piping, valves and fittings (modifies PG-58 and PG-59).

Science.gov (United States)

2010-10-01

... other suitable means employed to reduce the effects of metal temperature differentials. (e) Blowoff...-59). (a) Boiler external piping within the jurisdiction of the ASME Boiler and Pressure Vessel Code... and Pressure Vessel Code, boiler external piping must: (1) Meet the design conditions and criteria in...

SIMULATION OF NEGATIVE PRESSURE WAVE PROPAGATION IN WATER PIPE NETWORK

Directory of Open Access Journals (Sweden)

Tang Van Lam

2017-11-01

Full Text Available Subject: factors such as pipe wall roughness, mechanical properties of pipe materials, physical properties of water affect the pressure surge in the water supply pipes. These factors make it difficult to analyze the transient problem of pressure evolution using simple programming language, especially in the studies that consider only the magnitude of the positive pressure surge with the negative pressure phase being neglected. Research objectives: determine the magnitude of the negative pressure in the pipes on the experimental model. The propagation distance of the negative pressure wave will be simulated by the valve closure scenarios with the help of the HAMMER software and it is compared with an experimental model to verify the quality the results. Materials and methods: academic version of the Bentley HAMMER software is used to simulate the pressure surge wave propagation due to closure of the valve in water supply pipe network. The method of characteristics is used to solve the governing equations of transient process of pressure change in the pipeline. This method is implemented in the HAMMER software to calculate the pressure surge value in the pipes. Results: the method has been applied for water pipe networks of experimental model, the results show the affected area of negative pressure wave from valve closure and thereby we assess the largest negative pressure that may appear in water supply pipes. Conclusions: the experiment simulates the water pipe network with a consumption node for various valve closure scenarios to determine possibility of appearance of maximum negative pressure value in the pipes. Determination of these values in real-life network is relatively costly and time-consuming but nevertheless necessary for identification of the risk of pipe failure, and therefore, this paper proposes using the simulation model by the HAMMER software. Initial calibration of the model combined with the software simulation results and

Maximum discharge rate of liquid-vapor mixtures from vessels

International Nuclear Information System (INIS)

Moody, F.J.

1975-09-01

A discrepancy exists in theoretical predictions of the two-phase equilibrium discharge rate from pipes attached to vessels. Theory which predicts critical flow data in terms of pipe exit pressure and quality severely overpredicts flow rates in terms of vessel fluid properties. This study shows that the discrepancy is explained by the flow pattern. Due to decompression and flashing as fluid accelerates into the pipe entrance, the maximum discharge rate from a vessel is limited by choking of a homogeneous bubbly mixture. The mixture tends toward a slip flow pattern as it travels through the pipe, finally reaching a different choked condition at the pipe exit

Review of in-service thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1984-01-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper-shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (E 509-74) with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

The velocity of missiles generated by the disintegration of gas-pressurized vessels and pipes

International Nuclear Information System (INIS)

Baum, M.R.

1984-01-01

A theoretical model is developed to describe the velocity of fragments generated when a gas-pressurized vessel disintegrates. The predictions are compared with new and existing experimental data for spherical and cylindrical vessels and are shown to be an improvement over the widely used empirical correlation developed by Moore. It is also shown that, by an appropriate definition of the energy available for doing work on the fragments, the velocity of the fragments from the disintegration of a section of gas pipeline may be predicted by the same model

Assessment of water pipes durability under pressure surge

Science.gov (United States)

Pham Ha, Hai; Minh, Lanh Pham Thi; Tang Van, Lam; Bulgakov, Boris; Bazhenova, Soafia

2017-10-01

Surge phenomenon occurs on the pipeline by the closing valve or pump suddenly lost power. Due to the complexity of the water hammer simulation, previous researches have only considered water hammer on the single pipe or calculation of some positions on water pipe network, it have not been analysis for all of pipe on the water distribution systems. Simulation of water hammer due to closing valve on water distribution system and the influence level of pressure surge is evaluated at the defects on pipe. Water hammer on water supply pipe network are simulated by Water HAMMER software academic version and the capacity of defects are calculated by SINTAP. SINTAP developed from Brite-Euram projects in Brussels-Belgium with the aim to develop a process for assessing the integrity of the structure for the European industry. Based on the principle of mechanical fault, indicating the size of defects in materials affect the load capacity of the product in the course of work, the process has proposed setting up the diagram to fatigue assessment defect (FAD). The methods are applied for water pipe networks of Lien Chieu district, Da Nang city, Viet Nam, the results show the affected area of wave pressure by closing the valve and thereby assess the greatest pressure surge effect to corroded pipe. The SINTAP standard and finite element mesh analysis at the defect during the occurrence of pressure surge which will accurately assess the bearing capacity of the old pipes. This is one of the bases to predict the leakage locations on the water distribution systems. Amount of water hammer when identified on the water supply networks are decreasing due to local losses at the nodes as well as the friction with pipe wall, so this paper adequately simulate water hammer phenomena applying for actual water distribution systems. The research verified that pipe wall with defect is damaged under the pressure surge value.

Risk analysis of in-service pressure piping containing defects

International Nuclear Information System (INIS)

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

2004-01-01

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

Method and device for feeding purified water to a pressure vessel

International Nuclear Information System (INIS)

Hirato, Miharu.

1982-01-01

Purpose: To prevent thermal wear at the junction of feedwater pipes and purified water pipes, as well as maintain the function of the purified water feeding system by stopping the introduction of purified water to the heated water feeding system and introducing purified water to the recycling water system upon transient operation or start-up. Constitution: Since a feedwater heater does not function well during transient operation or upon start-up, the temperature of heated water flowing through the feedwater pipe is reduced to produce a temperature difference relative to the set temperature for the purified water feeding system. The temperature difference is detected by a temperature sensor and, when it arrives at a predetermined difference, an operation valve is switched to interrupt the feed of the purified water to the heated water feeding system and it is sent to a water recycling system. Then, the purified water is sent from the water recycling system by way of the discharge portion to the inside of a pressure vessel. Thus, since only the heated water flows to the junction between the cleaned water pipes and the heated water pipes, neither shocks are generated nor the performance of the purified water feeding system is impaired. (Moriyama, K.)

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

International Nuclear Information System (INIS)

Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

1990-10-01

This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns

Laguna Verde annulus pressurization loads evaluation

International Nuclear Information System (INIS)

Castaneda, M. A.; Cruz, M. A.; Cardenas, J. B.; Vargas, A.; Cruz, H. J.; Mercado, J. J.

2010-10-01

Annulus pressurization, jet impingement, pipe whip restraint and jet thrust are phenomena related to postulated pipe ruptures. A postulated pipe rupture at the weld between recirculation, or feedwater piping and a reactor nozzle safe end, will lead to a high flow rate of flashing water/steam mixture into the annulus between the reactor pressure vessel and the biological shield wall. The total effect of the vessel and pipe inventory blowdown from the break being postulated must be accounted for in the evaluation. A recirculation line break will give rise to an angular dependent short term pressure differential around the vessel, followed by a longer term pressure buildup in the annulus. A recirculation line postulated rupture may not produce worst case conditions and reference to time intervals for only the recirculation break should be treated superficially. A postulated rupture of the feedwater piping may produce the extreme case for determining: 1) the shield wall and reactor vessel to pedestal interactions, 2) loading on the reactor vessel internals, or 3) responses for the balance of piping attached to the vessel. Recently it was identified a potential issue regarding the criteria used to determine which cases were evaluated for Annulus Pressurization (A P) loads for new loads plants. The original A P loads methodology in the late 1970 and early 1980 years separated the mass/energy release calculation from the structural response calculation based on the implicit assumption that the maximum overall mass/energy release will result in maximizing the structural response and corresponding stresses on the reactor pressure vessel, internals, and containment structures. This process did not consider the dynamic response in the primary and secondary safety related structures, components and equipment. Consequently, the A P loads used as input for design adequacy evaluations of Nuclear Steam Supply System safety related components for new loads plants might have

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Pressure suppression apparatus of a nuclear power plant

International Nuclear Information System (INIS)

Mizumachi, W.; Funalashi, T.

1980-01-01

Pressure suppression apparatus for a nuclear reactor comprises a vessel surrounding a reactor pressure vessel and containing a water pool at the bottom of the vessel, and a steam exhaust pipe. The apparatus further comprises an exhaust chamber connected to the immersed portion of the exhaust pipe and provided with a number of discharge openings. (auth)

Device of connecting the metal sheet lining a concrete enclosure to a pipe opening inside the enclosure

International Nuclear Information System (INIS)

Petit, Guy.

1975-01-01

Said invention relates to a sealed device connecting a metal sheet anchored on the internal side of a concrete vessel containing a hot pressurized fluid, with a metallic pipe opening inside said vessel. It is intended for heat insulating structures so-called 'hot skin' used for the pressure vessels of some boiling water reactors. Said invention is intended for different types of said pipe such as: the penetrations for the inlets and outlets of the primary circuit, or anchoring cylindrical sheaths used as supports of components or other elements located inside said pressure vessel [fr

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

International Nuclear Information System (INIS)

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

2013-01-01

Accurate evaluation of the occurrence location and amplitude of pressure pulsations in piping systems can lead to efficient plant maintenance by preventing fatigue failure of piping and components because the pulsations can be one of the main causes of vibration fatigue and acoustic noise in piping. A non-destructive field method to measure pressure pulsations easily and directly was proposed to replace conventional methods such as prediction using numerical simulations and estimation using locally installed pressure gauges. The proposed method was validated experimentally by measuring pulsating flow in a mock-up piping system. As a result, it was demonstrated that the method to combine strain measurement on the outer surface of pipe with the formula for thick-walled cylinders could measure amplitudes and behavior of the pressure pulsations with a practical accuracy. Factors affecting the measurement accuracy of the proposed method were also discussed. Furthermore, the applicability of the formula for thin-walled cylinders was examined for variously shaped pipes. (author)

Sludge pipe flow pressure drop prediction using composite power ...

African Journals Online (AJOL)

Sludge pipe flow pressure drop prediction using composite power-law friction ... Water SA. Journal Home · ABOUT THIS JOURNAL · Advanced Search · Current Issue ... When predicting pressure gradients for the flow of sludges in pipes, the ...

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

International Nuclear Information System (INIS)

Rucinski, Russell A.

2002-01-01

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

High-performance fiber/epoxy composite pressure vessels

Science.gov (United States)

Chiao, T. T.; Hamstad, M. A.; Jessop, E. S.; Toland, R. H.

1978-01-01

Activities described include: (1) determining the applicability of an ultrahigh-strength graphite fiber to composite pressure vessels; (2) defining the fatigue performance of thin-titanium-lined, high-strength graphite/epoxy pressure vessel; (3) selecting epoxy resin systems suitable for filament winding; (4) studying the fatigue life potential of Kevlar 49/epoxy pressure vessels; and (5) developing polymer liners for composite pressure vessels. Kevlar 49/epoxy and graphite fiber/epoxy pressure vessels, 10.2 cm in diameter, some with aluminum liners and some with alternation layers of rubber and polymer were fabricated. To determine liner performance, vessels were subjected to gas permeation tests, fatigue cycling, and burst tests, measuring composite performance, fatigue life, and leak rates. Both the metal and the rubber/polymer liner performed well. Proportionately larger pressure vessels (20.3 and 38 cm in diameter) were made and subjected to the same tests. In these larger vessels, line leakage problems with both liners developed the causes of the leaks were identified and some solutions to such liner problems are recommended.

Effect of combined loading due to bending and internal pressure on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Sakai, Shinsuke

2006-01-01

Considering a rational maintenance rule of Light Water Reactor piping, reliable flaw evaluation criteria are essential to determine how a detected flaw is detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes to be considered for carbon steel piping, and can be analyzed by the elastic-plastic fracture mechanics. Some analytical efforts have been provided as flaw evaluation criteria using load correction factors such like the Z-factors in the JSME codes on fitness-for-service for nuclear power plants or the ASME boiler and pressure vessel code section XI. The present correction factors were conventionally determined taken conservatism and simplicity into account, however, the effect of internal pressure which would be an important factor under an actual plant condition was not adequately considered. Recently, a J-estimation scheme, 'LBB. ENGC' for ductile fracture analysis of circumferentially through-wall-cracked pipes subjected combined loading was newly developed to have a better prediction with more realistic manner. This method is explicitly incorporated the contribution of both bending and tension due to internal pressure by means of the scheme compatible with an arbitrary combined loading history. In this paper, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. A correction factor based on the new J-estimation scheme was compared with the present correction factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of internal pressure. (author)

Pipe resource evaluation according to the standards of Eurocode prEN13445-3 and Russian SNiP 2.05.06-85

International Nuclear Information System (INIS)

Jutas, A.; Kersiene, N.; Leisis, V.; Ziliukas, A.

2003-01-01

The use of different standards for the pressure vessels design educes some phenomena what by different methodologies are estimated. In Lithuania the newly found interest is driven by the desire to develop the design methodologies that are fundamentally based on principles of pressure vessels design and resource evaluation. In this article the differences of the Eurocode prEN13445-3 and Russian SNiP 2.05.06-85 and stress analysis of the pipes is shown too. These standards designed for calculation and assessment of exploitation loads of new pressure vessels. Usually corroded pipes are replaced then new and old pipe together welded. These pipelines for the pressure tests are very sensitive and re-calculation of limited pressure with assessment of geometry' changes of the corroded pipes is needful. (author)

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

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

Flexible Composite-Material Pressure Vessel

Science.gov (United States)

Brown, Glen; Haggard, Roy; Harris, Paul A.

2003-01-01

A proposed lightweight pressure vessel would be made of a composite of high-tenacity continuous fibers and a flexible matrix material. The flexibility of this pressure vessel would render it (1) compactly stowable for transport and (2) more able to withstand impacts, relative to lightweight pressure vessels made of rigid composite materials. The vessel would be designed as a structural shell wherein the fibers would be predominantly bias-oriented, the orientations being optimized to make the fibers bear the tensile loads in the structure. Such efficient use of tension-bearing fibers would minimize or eliminate the need for stitching and fill (weft) fibers for strength. The vessel could be fabricated by techniques adapted from filament winding of prior composite-material vessels, perhaps in conjunction with the use of dry film adhesives. In addition to the high-bias main-body substructure described above, the vessel would include a low-bias end substructure to complete coverage and react peak loads. Axial elements would be overlaid to contain damage and to control fiber orientation around side openings. Fiber ring structures would be used as interfaces for connection to ancillary hardware.

Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

International Nuclear Information System (INIS)

Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

2012-01-01

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

How simulation of failure risk can improve structural reliability - application to pressurized components and pipes

OpenAIRE

Cioclov, Dimitru Dragos

2013-01-01

Probabilistic methods for failure risk assessment are introduced, with reference to load carrying structures, such as pressure vessels (PV) and components of pipes systems. The definition of the failure risk associated with structural integrity is made in the context of the general approach to structural reliability. Sources of risk are summarily outlined with emphasis on variability and uncertainties (V&U) which might be encountered in the analysis. To highlight the problem, in its practical...

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

Prestressed cast iron pressure vessels as burst-proof pressure vessels for innovative nuclear applications

International Nuclear Information System (INIS)

Froehling, W.; Boettcher, A.; Bounin, D.; Steinwarz, W.; Geiss, M.; Trauth, M.

2000-01-01

The amendment to the German Atomic Energy Act from July 28, 1994 requires that events 'whose occurrence is practically excluded by the measures against damages', i.e. events of the category residual risk, must not necessitate far reaching protective measures outside the plant. For a conventional reactor pressure vessel, the residual risk consists in the very small probability of a catastrophic failure (formation of a large fracture opening, bursting of the vessel). With a prestressed cast iron vessel (PCIV), the formation of a large fracture opening or bursting of the vessel, respectively, is impossible due to its design properties. Against this background the possibility of the use of this type of pressure vessel for lightwater reactors has been studied in the frame of a 'Working Group for Innovative Nuclear Technology', founded by different research institutes and industrial companies. Furthermore, it has been studied whether the use of the PCIV support the realization of a corecatcher system. The results are presented in this report. Already many years earlier, Siempelkamp has performed industrial development and Forschungszentrum Juelich related experimental and theoretical safety research for the PCIV as an innovative, bust-proof pressure vessel concept. This development of the PCIV as well as its safety properties are also presented in a conclusive manner. (orig.) [de

The influence of gouge defects on failure pressure of steel pipes

International Nuclear Information System (INIS)

Alang, N A; Razak, N A; Zulfadli, M R

2013-01-01

Failure pressure of API X42 steel pipes with gouge defects was estimated through a nonlinear finite element (FE) analysis. The effect of gouge length on failure pressure of different pipe diameters was investigated. Stress modified critical strain (SMCS) model was applied as in predicting the failure of the pipe. The model uses strain based criteria to predict the failure. For validation of the model, the FE results were compared to experimental data in literature showing overall good agreement. The results show that the gouge length has significant influence on failure pressure. A smaller pipe diameter gives highest value of failure pressure

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.)

Model tests for prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Stoever, R.

1975-01-01

Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

Reactor pressure vessel design

International Nuclear Information System (INIS)

Foehl, J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

Effect of combined loading due to bending and internal pressure on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Sakai, Shinsuke

2008-01-01

Considering a rule for the rationalization of maintenance of Light Water Reactor piping, reliable flaw evaluation criteria are essential for determining how a detected flaw will be detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes that must be considered for carbon steel piping and can be analyzed by elastic-plastic fracture mechanics. Some analytical efforts have provided various flaw evaluation criteria using load correction factors, such as the Z-factors in the JSME codes on fitness-for-service for nuclear power plants and the section XI of the ASME boiler and pressure vessel code. The present Z-factors were conventionally determined, taking conservativity and simplicity into account; however, the effect of internal pressure, which is an important factor under actual plant conditions, was not adequately considered. Recently, a J-estimation scheme, LBB.ENGC for the ductile fracture analysis of circumferentially through-wall-cracked pipes subjected to combined loading was developed for more accurate prediction under more realistic conditions. This method explicitly incorporates the contributions of both bending and tension due to internal pressure by means of a scheme that is compatible with an arbitrary combined-loading history. In this study, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. The Z-factor obtained in this study was compared with the presently used Z-factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of the internal pressure. (author)

Proactive life extension of pressure vessels

Science.gov (United States)

Mager, Lloyd

1998-03-01

For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

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

Device for positioning ultrasonic probes and/or television cameras on the outer surface of reactor pressure vessels

International Nuclear Information System (INIS)

Zipser, R.; Dose, G.F.

1977-01-01

The device makes possible periodical in-service inspections of welding seams and material of a reactor pressure vessel without local human presence. A 'support ring' encloses the pressure vessel in a horizontal plane with free space. It is vertically moved up and down in the space between pressure vessel and thermal shield by means of tackles. At a control desk placed in a protected area its movement is controlled and its vertical position is indicated. A 'rotating track' with its own drive is rotating remote-controlled on the 'support ring'. By a combination of the vertical with the rotating movement, an ultrasonic probe placed removably on the 'rotating hack', or a television camera will be brought to any position on the cylindrical circumference of the pressure vessel. Special devices extend the radius of action, in upward direction for inspecting the welding seams of the coolant nozzles, and in downward direction for the inspection of welds on the hemispherical bottom of the pressure vessel or on the outlet pipe nozzle placed there. The device remains installed during reactor operation, but is moved down to the lower horizontal surface of the thermal shield. Parts which are sensible to radiation like probes or television cameras and special devices will then be removed respectively mounted before beginning an inspection compaign. This position may be reached by the lower access in the biological shield and through an opening in the horizontal surface of the thermal shield. (HP) [de

Design criteria for piping and nozzles program

International Nuclear Information System (INIS)

Moore, S.E.; Bryson, J.W.

1977-01-01

This report reviews the activities and accomplishments of the Design Criteria for Piping and Nozzles program being conducted by the Oak Ridge National Laboratory for the period July 1, 1975, to September 30, 1976. The objectives of the program are to conduct integrated experimental and analytical stress analysis studies of piping system components and isolated and closely-spaced pressure vessel nozzles in order to confirm and/or improve the adequacy of structural design criteria and analytical methods used to assure the safe design of nuclear power plants. Activities this year included the development of a finite-element program for analyzing two closely spaced nozzles in a cylindrical pressure vessel; a limited-parameter study of vessels with isolated nozzles, finite-element studies of piping elbows, a fatigue test of an out-of-round elbow, summary and evaluation of experimental studies on the elastic-response and fatigue failure of tees, parameter studies on the behavior of flanged joints, publication of fifteen topical reports and papers on various experimental and analytical studies; and the development and acceptance of a number of design rules changes to the ASME Code. 2 figures, 2 tables

Underwater cutting of stainless steel plate and pipe for dismantling reactor pressure vessels

International Nuclear Information System (INIS)

Hamasaki, M.; Tateiwa, F.; Kanatani, F.; Yamashita, S.

1982-01-01

A consumable electrode water jet cutting technique is described. Satisfactory underwater cutting of 80mm stainless steel plate using a current of 2000A and at a water depth of 200mm has been demonstrated. The electrical requirements for this arc welding method applied to cutting were found to be approximately one third those required for conventional plasma arc cutting for the same thickness plate. An application of this technique might be found in the dismantling of atomic reactor pressure vessels, and parts of commercial atomic reactors. (author)

Pressure test method for reactor pressure vessel in construction field

International Nuclear Information System (INIS)

Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

1998-01-01

Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

Metallurgy of steels for PWR pressure vessels

International Nuclear Information System (INIS)

Kepka, M.; Mocek, J.; Barackova, L.

1980-01-01

A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

Metallurgy of steels for PWR pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

1980-09-01

A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

DETERMINATION OF LIQUID FILM THICKNESS FOLLOWING DRAINING OF CONTACTORS, VESSELS, AND PIPES IN THE MCU PROCESS

International Nuclear Information System (INIS)

Poirier, M; Fernando Fondeur, F; Samuel Fink, S

2006-01-01

The Department of Energy (DOE) identified the caustic side solvent extraction (CSSX) process as the preferred technology to remove cesium from radioactive waste solutions at the Savannah River Site (SRS). As a result, Washington Savannah River Company (WSRC) began designing and building a Modular CSSX Unit (MCU) in the SRS tank farm to process liquid waste for an interim period until the Salt Waste Processing Facility (SWPF) begins operations. Both the solvent and the strip effluent streams could contain high concentrations of cesium which must be removed from the contactors, process tanks, and piping prior to performing contactor maintenance. When these vessels are drained, thin films or drops will remain on the equipment walls. Following draining, the vessels will be flushed with water and drained to remove the flush water. The draining reduces the cesium concentration in the vessels by reducing the volume of cesium-containing material. The flushing, and subsequent draining, reduces the cesium in the vessels by diluting the cesium that remains in the film or drops on the vessel walls. MCU personnel requested that Savannah River National Laboratory (SRNL) researchers conduct a literature search to identify models to calculate the thickness of the liquid films remaining in the contactors, process tanks, and piping following draining of salt solution, solvent, and strip solution. The conclusions from this work are: (1) The predicted film thickness of the strip effluent is 0.010 mm on vertical walls, 0.57 mm on horizontal walls and 0.081 mm in horizontal pipes. (2) The predicted film thickness of the salt solution is 0.015 mm on vertical walls, 0.74 mm on horizontal walls, and 0.106 mm in horizontal pipes. (3) The predicted film thickness of the solvent is 0.022 mm on vertical walls, 0.91 mm on horizontal walls, and 0.13 mm in horizontal pipes. (4) The calculated film volume following draining is: (a) Salt solution receipt tank--1.6 gallons; (b) Salt solution feed

Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

International Nuclear Information System (INIS)

Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

1975-01-01

Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

Lateral impact on pressurized glass/epoxy pipes

Energy Technology Data Exchange (ETDEWEB)

Santiago, Rafael C.; Alves, Marcilio [Universidade de Sao Paulo (USP), SP (Brazil). Escola Politecnica], e-mail: rafael.santiago@poli.usp.br, e-mail: maralves@usp.br

2009-07-01

This work studies the impact of a drop mass against glass/epoxy composite pipes. The experimental tests comprises an impact hammer and the set-up is such that a drop mass falls transversally to the main length of the tube. The impact response is measured with a non-contact laser system and reveals some issues explored here in a numerical model. Inner pressure was applied to the tube in order to access its influence on failure. It has been found that the numerical model represents well the structural collapse of a pressurized pipe under a lateral impact. (author)

Estimation on the Flow Phenomena and the Pressure Loss for the Inlet Part of a Research Reactor Vessel

International Nuclear Information System (INIS)

Seo, Kyoung Woo; Oh, Jae Min; Seo, Jae Kwang; Yoon, Ju Hyeon; Lee, Doo Jeong

2009-01-01

For a research reactor, a conceptual primary cooling system (PCS) was designed for an adequate cooling to the reactor core. The developed primary cooling circuit consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. The main function of the primary cooling pumps (PCPs) of the PCS was to circulate the reactor coolant through the fuel core and the heat exchangers during a normal operation. The head according to the design flow rate which was determined by the thermal hydraulic design analysis for the core should be estimated to design the PCPs in the fluid system. The pressure loss in the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. However, it is insufficient to estimate the pressure loss for 3-dimensional flow phenomena such as the flow path in the reactor with the theoretical dimensional analysis based on experimental data. The purpose of this research is to evaluate the pressure loss of the part of a research reactor vessel. For evaluating the pressure loss, the commercially available CFD computer model, FLUENT, was employed. First, for validating the application of FLUENT to the pressure loss, a simple case was calculated and compared with the Idelchik empirical correlation. Secondly, several cases for the inlet part of a research reactor vessel were estimated by a FLUENT 3- dimensional calculation

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

International Nuclear Information System (INIS)

Estrade, J.

1985-01-01

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

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

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

International Nuclear Information System (INIS)

Scherner, D.

1985-01-01

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

Some aspects of reactor pressure vessel integrity

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic, G.J.

1996-01-01

Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

International Nuclear Information System (INIS)

Ventura, Mirta A.; Rosso, Ricardo D.

2002-01-01

In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

78 FR 45271 - Welded Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam

Science.gov (United States)

2013-07-26

... Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam Determination On the basis of the record... reason of imports from Malaysia, Thailand, and Vietnam of welded stainless steel pressure pipe, provided... contained in USITC Publication 4413 (July 2013), entitled Welded Stainless Steel Pressure Pipe from Malaysia...

Ratcheting of pressurized piping subjected to seismic loading

International Nuclear Information System (INIS)

Scavuzzo, R.J.; Lam, P.C.; Gau, J.S.

1992-01-01

The ABAQUS finite element code was used to model a pressurized pipe and subjected to cyclic bending loads to investigate ratcheting. A 1-in. schedule 40 pipe was loaded with a slow (static) cyclic load. The pipe internal pressure was varied from 0 to 6000 psi. In this paper, two types of materials were considered: an elastic perfectly plastic and a bilinear elastic-plastic material. Two types of finite elements of the ABAQUS program were compared to analytical solutions to evaluate the element accuracy in the plastic regime. Depending upon loading conditions and specified material properties, three different responses were observed from the finite element analyses: cyclic plasticity, ratcheting of the hoop strain, or shakedown. These analytical results are compared to some experimental measurements

PWR pressurizer discharge piping system on-site testing

International Nuclear Information System (INIS)

Anglaret, G.; Lasne, M.

1983-08-01

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

Pressure releasing device for reactor container

International Nuclear Information System (INIS)

Takeda, Mika.

1994-01-01

In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Foundamental characteristics of layered pressure vessel

International Nuclear Information System (INIS)

Moriwaki, Yoshikazu; Fugino, Masayuki; Shimizu, Yasuhiro; Nakamura, Takeshi

1978-01-01

Pressure vessels become larger and the working pressure become higher with the remarkable development of petroleum, chemical, thermal power generation and atomic energy industries. Multi-layered pressure vessels can be manufactured cheaply without large installations, and large wall thickness can be made, therefore they are suitable for large pressure vessels. The stress and deformation behaviors of such vessels are very complex because of the effect of frictional force working between layers. In this study, the phenomena arising in multiple layers and the difference as compared with single wall were studied fundamentally as one step for analyzing multi-layered pressure vessels as a whole. Finite element technique was employed as the analyzing method, and the behavior of multiple layers was analyzed, regarding it as multiple contact problem. The behavior of multiple layers seems to appear conspicuously in case of bending load, therefore the basic characteristics regarding bending were examined. The evaluation of interfacial stiffness was carried out by experiment. The computer program for analyzing multiple contact problem was developed. In order to examine the validity of the program, comparison with the analytical solution heretofore and the result of calculation by finite element technique was carried out. Moreover, the experimental proof with multi-layered models was made. The frictional force between layers hardly contributes to the stiffness. (Kako, I.)

46 CFR 115.812 - Pressure vessels and boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter. [CGD 85-080, 61...

Conformable pressure vessel for high pressure gas storage

Science.gov (United States)

Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

2016-01-12

A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

Pressure pulsation measurements in pipe and cluster flows

International Nuclear Information System (INIS)

Benemann, A.; Voj, P.

1976-01-01

Measuring and evaluation techniques of pressure pulsations in pipe and cluster flows are described. The measurements were made on a 1 m long SNR rod-cluster and its feed and drain pipes. At Reynolds numbers in the cluster of 8.9 x 10 4 flow velocities of 14 m/sec were achieved. With the aid of a block diagram recording of the measured values by piezoelectric crystal and piezo-resistive strain gange as well as data processing are explained. For the analytical treatment of the pressure pulsation signals characterizing the turbulence field computer codes of a digital computer and a fast-fourier analyzer (Hewlett-Packard 5450 A) were used. The results show good agreement with theoretical curves on the behaviour of turbulent boundary layers of cluster and pipe flows at high Reynolds numbers. (TK) [de

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

Guidelines for pressure vessel safety assessment

Science.gov (United States)

Yukawa, S.

1990-04-01

A technical overview and information on metallic pressure containment vessels and tanks is given. The intent is to provide Occupational Safety and Health Administration (OSHA) personnel and other persons with information to assist in the evaluation of the safety of operating pressure vessels and low pressure storage tanks. The scope is limited to general industrial application vessels and tanks constructed of carbon or low alloy steels and used at temperatures between -75 and 315 C (-100 and 600 F). Information on design codes, materials, fabrication processes, inspection and testing applicable to the vessels and tanks are presented. The majority of the vessels and tanks are made to the rules and requirements of ASME Code Section VIII or API Standard 620. The causes of deterioration and damage in operation are described and methods and capabilities of detecting serious damage and cracking are discussed. Guidelines and recommendations formulated by various groups to inspect for the damages being found and to mitigate the causes and effects of the problems are presented.

Ductile failure of pipes with defects under combined pressure and bending

International Nuclear Information System (INIS)

Darlaston, B.J.L.; Harrison, R.P.

1977-01-01

The main part of the experimental programme was carried out on 3.5'' diam. pipes with a wall thickness of 0.064''. Various lengths of defect were assessed but only two depths, 0.044'' and 0.060''. Some full penetration defect tests were carried out under bending loading. The defects were 0.012'' wide and nominally flat bottomed. The tensile properties of the pipes were determined by taking specimens from each of the tubes. The pipes were exposed to pressure only test, bending only test and combined bending and pressure test. The results are given in tables. The observations led to the postulation of a design rule relating to the effect of defect in pipes under combined internal pressure and bending. It applies only to ductile situations in which the mode of failure is by a collapse mechanism: If the failure of a pipe containing an axial defect occurs by plastic collapse then provided the bending moment does not exceed half that for collapse due to bending alone, it will have a negligible effect on the failure pressure. (J.B.)

Investigation and evaluation of stress-corrosion cracking in piping of light water reactor plants

International Nuclear Information System (INIS)

1979-01-01

In 1975, a Pipe Cracking Study Group, established by the United States Nuclear Regulatory Commission (USNRC), reviewed intergranular stress-corrosion cracking (IGSCC) in Bioling Water Reactors (BWRs) and issued a report. During 1978, IGSCC was reported for the first time in large-diameter piping (> 20 in.) in a BWR in Germany. This discovery, together with the reported questions concerning the interpretation of ultrasonic inspections, led to the activation of a new Pipe Crack Study Group (PCSG) by USNRC. The charter of the new PCSG was expanded: (1) to include review of potential for stress-corrosion cracking in Pressurized Water Reactors (PWRs) as well as BWRs, (2) to examine operating experience in foreign reactors relevant to IGSCC, and (3) to study five specific questions. The PCSG limited the scope of the study to BWR and PWR piping runs and safe ends attached to the reactor pressure vessel. Not considered were components such as the reactor pressure vessel, pumps, valves, steam generators, large steam turbines, etc. Throughout this report, as well as in the title, the safe ends are arbitrarily defined as piping

Control of the ORR-PSF pressure-vessel surveillance irradiation experiment temperature

International Nuclear Information System (INIS)

Miller, L.F.

1982-01-01

Control of the Oak Ridge Research Reactor Pool Side Facility (ORR-PSF) pressure vessel surveillance irradiation experiment temperature is implemented by digital computer control of electrical heaters under fixed cooling conditions. Cooling is accomplished with continuous flows of water in pipes between specimen sets and of helium-neon gas in the specimen set housings. Control laws are obtained from solutions of the discrete-time Riccati equation and are implemented with direct digital control of solid state relays in the electrical heater circuit. Power dissipated by the heaters is determined by variac settings and the percent of time that the solid state relays allow power to be supplied to the heaters. Control demands are updated every forty seconds

Review of the models and mechanisms for environmentally-assisted crack growth of pressure vessel and piping steels in PWR environments

International Nuclear Information System (INIS)

Cullen, W.; Gabetta, G.; Hanninen, H.

1985-12-01

The crack-tip micromechanisms and the computational models for environmentally-assisted cracking in pressure vessel and piping steels in high-temperature, low-oxygen (PWR), reactor-grade water are described and evaluated in this report. The report begins with a brief description of the critical variables which are known to affect environmentally-assisted subcritical cracking in these metal/environment systems. The micromechanistic models are discussed in some detail, with anodic dissolution and hydrogen assistance being the prime candidates for the successful explanation of the observed phenomena. The anodic dissolution model offers far better quantification of the environmentally-assisted crack growth rates, but tends to overpredict the rates for a large number of conditions. The hydrogen assistance models qualitatively could account for a wider range of effects, but quantification of the model is virtually nonexistent. A variety of calculational models are in various stages of development; all of them are far from use as a predictive tool. Crack-tip strain rate models have received the most attention, and the approach to their use has been to partition the environmentally-assisted growth rates into a mechanically-driven component, with the environmental enhancement superposed. The environment component is then correlated with a calculated crack-tip strain rate. 141 refs., 59 figs

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).

How to replace a reactor pressure vessel

International Nuclear Information System (INIS)

Huber, R.

1996-01-01

A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

The pressure vessel for the NSF tandem

International Nuclear Information System (INIS)

Jones, C.W.

1979-04-01

The pressure vessel is a major component of the 30 MV tandem Van de Graaff electrostatic accelerator to be used in nuclear structure research at Daresbury Laboratory. The accelerator will be capable of accelerating the full range of ions in the form of a beam. Acceleration takes place in a vertical evacuated tube (beam tube) by means of a high potential on a terminal at the central position, the terminal and beam tube assembly being supported by an insulated stack structure within the pressure vessel. Under operating conditions the vessel is filled with sulphur hexafluoride gas (SF 6 ) at high pressure which acts as an insulating medium between the centre terminal and the vessel wall. The vessel is situated inside a concrete tower which besides supporting the injector room above the vessel also acts as radiation shielding around the accelerator. The report covers: functional requirements; fundamental considerations with regard to the design and procurement; detail design; materials; manufacture; acceptance test; surface treatment; final leak test. (U.K.)

Head spray nozzle in reactor pressure vessel

International Nuclear Information System (INIS)

Hatano, Shun-ichi.

1990-01-01

In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

An Overview of Corroded Pipe Repair Techniques Using Composite Materials

OpenAIRE

K. S. Lim; S. N. A. Azraai; N. M. Noor; N. Yahaya

2015-01-01

Polymeric composites are being increasingly used as repair material for repairing critical infrastructures such as building, bridge, pressure vessel, piping and pipeline. Technique in repairing damaged pipes is one of the major concerns of pipeline owners. Considerable researches have been carried out on the repair of corroded pipes using composite materials. This article attempts a short review of the subject matter to provide insight into various techniques used in repa...

Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

International Nuclear Information System (INIS)

Sato, I.; Suzuki, K.

1994-01-01

For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

The effect of a self-balancing through wall residual stress distribution on the extension of a through-wall crack in a pressure vessel

International Nuclear Information System (INIS)

Smith, E.

1993-01-01

Leak-before-break arguments for pressurized components involve a comparison of the critical size of crack that will grow unstably under accident loadings and the critical leakage crack size for normal operation loadings. The paper is concerned with the former crack size and particularly with regard to the effect of residual stresses on the critical unstable crack size. Results from an analysis of a simple simulation model are used to provide underpinning for the view, expressed by Green and Knowles at the 1992 American Society of Mechanical Engineers Pressure Vessel and Piping Conference, that self-balancing through-wall residual stresses have little overall effect on the extension of a through-wall crack in a pressure vessel

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Results of reactor pressure vessels ISI

International Nuclear Information System (INIS)

Cepcek, S.

1994-01-01

To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

Study on criterion for leak before break assessment of pressure pipes

International Nuclear Information System (INIS)

Yang Linjuan

2009-01-01

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

Holographic and acoustic emission evaluation of pressure vessels

International Nuclear Information System (INIS)

Boyd, D.M.

1980-01-01

Optical holographic interfereometry and acoustic emission monitoring were simultaneously used to evaluate two small, high pressure vessels during pressurization. The techniques provide pressure vessel designers with both quantitative information such as displacement/strain measurements and qualitative information such as flaw detection. The data from the holographic interferograms were analyzed for strain profiles. The acoustic emission signals were monitored for crack growth and vessel quality

Reactor pressure vessel status report

International Nuclear Information System (INIS)

Strosnider, J.; Wichman, K.; Elliot, B.

1994-12-01

This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

Study on acoustic emission signals of active defect in pressure piping under hydraulic pressure

International Nuclear Information System (INIS)

Ai Qiong; Liu Caixue; Wang Yao; He Pan; Song Jian

2009-01-01

Experimental investigations of acoustic emission (AE) of active defect in pressure piping with a prefabricated crack under hydraulic pressure tester were conducted. AE signals of fatigue-crack-growth in pressure piping were monitored incessantly in all processes, and all signals recorded were analyzed and processed. The result of signal processing show that the amplitude and energy of acoustic emission signals from defect in pressure pipeline increase gradually with the load time, and thus the active defects in pipeline can be identified; the amplitude, energy and count of acoustic emission signals increase sharply before the defect runs through, and we can forecast the penetrated leakage of pipeline. (authors)

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

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

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

International Nuclear Information System (INIS)

Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

2008-01-01

The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

International Nuclear Information System (INIS)

Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

1979-01-01

A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

Rethinking ASME III seismic analysis for piping operability evaluations

International Nuclear Information System (INIS)

Adams, T.M.; Stevenson, J.D.

1994-01-01

It has been recognized since the mid 1980's that there are very large seismic margins to failure for nuclear piping systems when designed using current industry practice, design criteria, and methods. As a result of this realization there are or have been approximately eighteen initiatives within the ASME , Boiler and Pressure Vessel Code Section III, Division 1, in the form of proposed code cases and proposed code text changes designed to reduce these failure margins to more realistic values. For the most part these initiatives have concentrated on reclassifying seismic inertia stresses in the piping as secondary and increasing the allowable stress limits permitted by Section III of the ASME, Boiler Code. This paper focuses on the application of non-linear spectral analysis methods as a method to reduce the input seismic demand determination and thereby reduce the seismic failure margins. The approach is evaluated using the ASME Boiler Pressure Vessel Code Section III Subgroup on Design benchmark procedure as proposed by the Subgroup's Special Task Group on Integrated Piping Criteria. Using this procedure, criteria are compared to current code criterion and analysis methods, and several other of the currently proposed Boiler and Pressure Vessel, Section III, changes. Finally, the applicability of the non-linear spectral analysis to continued Safe Operation Evaluations is reviewed and discussed

Pressure vessel integrity and weld inspection procedure

International Nuclear Information System (INIS)

Solomon, K.A.; Okrent, D.; Kastenberg, W.E.

1975-01-01

The primary objective of this paper is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an inter-relation between pressure vessel integrity, and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. A modified Markov process is employed and a computer code was written to obtain numerical results. The Markov process mathematically describes the following physical events. In a nuclear reactor pressure vessel weld, some defects may exist prior to the zeroth inspection (i.e., prior to vessel operation). During the zeroth inspection and repair processes, some of these defects are removed. During the first cycle of vessel operation, the existing defects may grow and some new defects may be generated. Those defects that are found at the first (and succeeding) inspection interval and warrant repair, are repaired. The above process continues through several operating cycles to the end of vessel life. During any inspection, only a portion of the welds may be inspected, and with less than perfect efficiency

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

International Nuclear Information System (INIS)

Mc Donald, B.N.

1976-01-01

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

Reactor pressure vessel support

International Nuclear Information System (INIS)

Butti, J.P.

1977-01-01

A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

Crack propagation on spherical pressure vessels

International Nuclear Information System (INIS)

Lebey, J.; Roche, R.

1975-01-01

The risk presented by a crack on a pressure vessel built with a ductile steel cannot be well evaluated by simple application of the rules of Linear Elastic Fracture Mechanics, which only apply to brittle materials. Tests were carried out on spherical vessels of three different scales built with the same steel. Cracks of different length were machined through the vessel wall. From the results obtained, crack initiation stress (beginning of stable propagation) and instable propagation stress may be plotted against the lengths of these cracks. For small and medium size, subject to ductile fracture, the resulting curves are identical, and may be used for ductile fracture prediction. Brittle rupture was observed on larger vessels and crack propagation occurred at lower stress level. Preceedings curves are not usable for fracture analysis. Ultimate pressure can be computed with a good accuracy by using equivalent energy toughness, Ksub(1cd), characteristic of the metal plates. Satisfactory measurements have been obtained on thin samples. The risks of brittle fracture may then judged by comparing Ksub(1cd) with the calculated K 1 value, in which corrections for vessel shape are taken into account. It is thus possible to establish the bursting pressure of cracked spherical vessels, with the help of two rules, one for brittle fracture, the other for ductile instability. A practical method is proposed on the basis of the work reported here

46 CFR 32.50-20 - Cargo piping for tank vessels constructed between November 10, 1936, and July 1, 1951-TB/ALL.

Science.gov (United States)

2010-10-01

... HOMELAND SECURITY TANK VESSELS SPECIAL EQUIPMENT, MACHINERY, AND HULL REQUIREMENTS Pumps, Piping, and Hose... may be omitted where the pipe is extra heavy, all joints are welded, and bends are installed to provide for expansion and contraction. (b) Cargo piping shall not pass through spaces containing machinery...

Computing the effect of plastic deformation of piping on pressure transient propagation

International Nuclear Information System (INIS)

Youngdahl, C.K.; Kot, C.A.

1977-01-01

The computer program PTA-1 performs pressure-transient analysis of large piping networks using the one-dimensional method of characteristics applied to a fluid-hammer formulation. The effect of elastic-plastic deformation of piping on pulse propagation is included in the computation. Each pipe is modeled as a series of rings, neglecting axial effects, bending moments, and inertia. The fluid wave speed is a function of pipe deformation and, consequently, of position and time. Comparison with existing experimental data indicate that this simple fluid-structure interaction model gives suprisingly accurate results for both pressure histories in the fluid and strain histories in the piping

Nickel hydrogen common pressure vessel battery development

Science.gov (United States)

Jones, Kenneth R.; Zagrodnik, Jeffrey P.

1992-01-01

Our present design for a common pressure vessel (CPV) battery, a nickel hydrogen battery system to combine all of the cells into a common pressure vessel, uses an open disk which allows the cell to be set into a shallow cavity; subsequent cells are stacked on each other with the total number based on the battery voltage required. This approach not only eliminates the assembly error threat, but also more readily assures equal contact pressure to the heat fin between each cell, which further assures balanced heat transfer. These heat fin dishes with their appropriate cell stacks are held together with tie bars which in turn are connected to the pressure vessel weld rings at each end of the tube.

Pressurized wet digestion in open vessels (T11)

International Nuclear Information System (INIS)

Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

2002-01-01

Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

A thermal insulation system intended for a prestressed concrete vessel

International Nuclear Information System (INIS)

Aubert, Gilles; Petit, Guy.

1975-01-01

The description is given of a thermal insulation system withstanding the pressure of a vaporisable fluid for a prestressed concrete vessel, particularly the vessel of a boiling water nuclear reactor. The ring in the lower part of the vessel has, between the fluid inlet pipes and the bottom of the vessel, an annular opening of which the bottom edge is integral with an annular part rising inside the ring and parallel to it. This ring is hermetically connected to the bottom of the vessel and is coated with a metal lagging, at least facing the annular opening. This annular opening is made in the ring half-way up between the fluid inlet pipes and the bottom of the vessel. It is connected to the bottom of the vessel through the internal structure enveloping the reactor core [fr

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

Pressure vessel rupture within a chamber: the pressure history on the chamber wall

International Nuclear Information System (INIS)

Baum, M.R.

1989-04-01

Generally there is a large number of pressure vessels containing high pressure gas on power stations and chemical plant. In many instances, particularly on power plant, these vessels are within the main building. If a pressure vessel were to fail, the surrounding structures would be exposed to blast loads and the forces resulting from jets of fluid issuing from the breached vessel. In the case where the vessel is in a relatively closed chamber there would also be a general overpressurisation of the chamber. At the design stage it is therefore essential to demonstrate that the plant could be safely shut down in the event of a pressure vessel failure, that is, it must be shown that the chamber will not collapse thus putting the building at risk or hazarding equipment essential for a safe shut down. Such an assessment requires the loads applied to the chamber walls, roof, etc. to be known. (author)

Procurement of replacement pressure vessels for MURR

International Nuclear Information System (INIS)

Meyer, W.A. Jr.; Edwards, C.B. Jr.; McKibben, J.C.; Schoone, A.R.

1989-01-01

The University of Missouri Research Reactor Facility (MURR) located in Columbia, Missouri, is the highest powered, highest steady-state flux university research reactor in the United States. The reactor is a 10-MW pressurized loop, in-pool-type, light-water-moderated, beryllium-reflected, flux trap reactor. MURR has a compact core (0.033 m 3 ) composed of eight fuel elements of the materials test reactor type arranged as an annular right circular cylinder between the inner and outer aluminum pressure vessels. Conservative engineering judgment resulted in the decision in 1988 to purchase new inner and outer pressure vessels. This paper details the difficulties encountered in procuring replacements for aluminum pressure vessels built to standards that are no longer applicable in attempting to meet nuclear standards that are not applicable to nonferrous material

Pressure wave propagation in the discharge piping with water pool

International Nuclear Information System (INIS)

Bang, Young S.; Seul, Kwang W.; Kim, In Goo

2004-01-01

Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined

Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

Directory of Open Access Journals (Sweden)

Hutli Ezddin

2017-01-01

Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

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

Nuclear reactor pressure vessel flaw distribution development

International Nuclear Information System (INIS)

Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

1991-12-01

Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

Environmental factor approach to account for water effects in pressure vessel and piping fatigue evaluations

International Nuclear Information System (INIS)

Mehta, H.S.; Gosselin, S.R.

1998-01-01

This paper summarizes past and current studies of the environmental fatigue effects in light water reactor (LWR) applications. Current Argonne and Japanese research efforts are reviewed and an approach to calculate an environmental correction factor is described. A description of how the proposed approach can be implemented in section III, NB-3600 and NB-3200-type fatigue evaluations is presented along with examples of applying the approach to piping (NB-3600) and safe end fatigue evaluations. These procedures were applied to several BWR and pressurized water reactor (PWR) example cases. The results of these case studies indicated that there is a modest increase in calculated fatigue usage, which is considerably less than the results obtained when the NUREG/CR-5999 curves are applied directly. (orig.)

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

Expanded Fermilab pressure vessel directory program

Energy Technology Data Exchange (ETDEWEB)

Tanner, A.

1983-01-01

Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

Expanded Fermilab pressure vessel directory program

International Nuclear Information System (INIS)

Tanner, A.

1983-01-01

Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

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

Metallurgical failure investigation of a pipe connector fracture of an expansion vessel; Werkstofftechnische Schadensuntersuchung des Abrisses einer Rohrverschraubung eines Ausgleichsbehaelters

Energy Technology Data Exchange (ETDEWEB)

Neidel, Andreas [Siemens AG, Power and Gas, Gas Turbines and Generators, Gasturbinenwerk Berlin (Germany). Werkstoffprueflabor

2016-08-15

A pipe connector of an expansion vessel of a safety heat exchanger was torn off in a test facility's natural gas compressor. From a material point of view, the cause of the damage is a fatigue fracture induced by pulsating bending stress. The fatigue fracture originated from both, the pipe's outer surface as well as from its inner surface, which is consistent with the given stress situation (pulsating bending stress). Material defects or welding-induced flaws were not observed. Corrosion, wear, or thermal overload which may have promoted the damage, were not observed either. The primary cause was a major design error. Cases of dynamic load were obviously not duly taken into account during designing, so that the free-swinging mass of the expansion vessel which was mounted to a pipe of a diameter of only half an inch and, furthermore, installed in an angle of 45 (additional static preload.), could cause the fatigue failure induced by pulsating bending stress in the zone of highest stresses at the transition of the expansion vessel and the the pipe connector due to dynamic operating loads which always occur in plants like these.

Feasibility of local stress relieving close to main shell of a large vessel

International Nuclear Information System (INIS)

Hancinsky, O.A.

1978-01-01

This work determines the feasibility of local stress relieving for a circumferential pipe-to-nozzle field weld positioned close to the main shell of a large pressure vessel. This is applicable to nuclear as well as conventional vessels. ANSYS computer program is utilized to perform thermal and thermal stress analysis and ASME Pressure Vessels Code is adhered to. Conclusions and recommendations are made with a view on their applicability in practice

Ultrasound periodic inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Haniger, L.

1980-01-01

Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

The design of lifting attachments for the erection of large diameter and heavy wall pressure vessels

International Nuclear Information System (INIS)

Antalffy, Leslie P.; Miller, George A.; Kirkpatrick, Kenneth D.; Rajguru, Anil; Zhu, Yong

2016-01-01

Lifting attachments for the erection of large diameter and heavy wall pressure vessels require special consideration to ensure that their attachment to their vessel shells or heads do not overstress the vessel during the erection process when lifting these from grade onto their respective foundations. Today, in refinery and petrochemical services, large diameter vessels with diameters ranging up to 15 m and reactors with lifting weights in the range of 700–1400 tons are not uncommon. In today's fabrication market, these vessels may be purchased and fabricated in shops dispersed globally and will require unique equipment for their safe handling, transportation and subsequent erection. The challenge is to design the lifting attachments in such a manner that the attachments provide a safe, cost effective and effective solution based upon the limitations of the job site lift equipment available for erection. Such equipment for the transportation and subsequent lifting of large diameter and heavy wall pressure equipment is usually scarce and quite expensive. Planning ahead, well in advance of the lift date is almost a mandatory requirement. Usually, the specific parameters of the vessel to be lifted and the lifting equipment available at the site will dictate the type of lifting attachments to be designed for the vessel. Once the type of vessel attachment has been chosen, careful consideration must be given to the design of attachments to the pressure vessel in consideration to ensure that the vessel and lifting components are not overstressed during the lifting process. The paper also discusses different types of lifting attachments that may be attached to each end of the vessel either by bolting or welding and discusses the pros and cons of each. The paper also provides an example of a finite element analysis (FEA) of a top nozzle, a FEA of a pair of lifting trunnions and a FEA of welded on lifting lugs for buried pipe. The purpose of the paper is to outline the

Creep deformation and crack growth in a low alloy steel welded pressure vessel containing defects

International Nuclear Information System (INIS)

Coleman, M.C.

1982-01-01

A full-size pressure vessel was tested for effects of welding residual stresses on creep deformation and crack growth. The vessel, based on 1/2 Cr 1/2 Mo 1/4 V main steam pipe, contained four 2CrMo manual metal arc welds, two in the as-welded condition and two stress-relieved. All the welds contained pre-existing defects machined in the heat affected zones. Testing was carried out at two internal steam pressures, 250 and 350 bar, and 565 0 C. Cracked and uncracked areas of the vessel were monitored continuously. Results are presented for the continuous creep deformation observed in both the hoop and axial directions of the welds throughout the 11,400 h of testing, as well as the intermittent strain data obtained during inspections. Crack growth observations are described based on nondestructive examination. The residual stresses measured are also given for both the as-welded and stress relieved weldments. Results obtained are discussed in terms of the effects of welding residual stress on the hoop and axial deformations observed in the welds. Similarly, the effects of residual stress on creep crack growth are considered together with compositional and microstructural implications. 9 figures, 5 tables

Development of PWR pressure vessel steels

International Nuclear Information System (INIS)

Druce, S.; Edwards, B.

1982-01-01

Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

Development of PWR pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Druce, S.; Edwards, B.

1982-01-01

Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Frictional pressure drop of gas liquid two-phase flow in pipes

International Nuclear Information System (INIS)

Shannak, Benbella A.

2008-01-01

Experiments of air water two-phase flow frictional pressure drop of vertical and horizontal smooth and relatively rough pipes were conducted, respectively. The result demonstrated that the frictional pressure drop increases with increasing relative roughness of the pipe. However, the influence of the relative roughness becomes more evident at higher vapour quality and higher mass flux. A new prediction model for frictional pressure drop of two-phase flow in pipes is proposed. The model includes a new definition of the Reynolds number and the friction factor of two-phase flow. The proposed model fits the presented experimental data very well, for vertical, horizontal, smooth and rough pipes. Therefore, the reproductive accuracy of the model is tested on the experimental data existing in the open literature and compared with the most common models. The statistical comparison, based on the Friedel's Data-Bank containing of about 16,000 measured data, demonstrated that the proposed model is the best overall agreement with the data. The model was tested for a wide range of flow types, fluid systems, physical properties and geometrical parameters, typically encountered in industrial piping systems. Hence, calculating based on the new approach is sufficiently accurate for engineering purposes

Pressurized Vessel Slurry Pumping

International Nuclear Information System (INIS)

Pound, C.R.

2001-01-01

This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

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

Analytical considerations in the code qualification of piping systems

International Nuclear Information System (INIS)

Antaki, G.A.

1995-01-01

The paper addresses several analytical topics in the design and qualification of piping systems which have a direct bearing on the prediction of stresses in the pipe and hence on the application of the equations of NB, NC and ND-3600 of the ASME Boiler and Pressure Vessel Code. For each of the analytical topics, the paper summarizes the current code requirements, if any, and the industry practice

Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

International Nuclear Information System (INIS)

Rodabaugh, E.C.

1983-06-01

The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

Transportable, small high-pressure preservation vessel for cells

International Nuclear Information System (INIS)

Kamimura, N; Sotome, S; Shimizu, A; Nakajima, K; Yoshimura, Y

2010-01-01

We have previously reported that the survival rate of astrocytes increases under high-pressure conditions at 4 0 C. However, pressure vessels generally have numerous problems for use in cell preservation and transportation: (1) they cannot be readily separated from the pressurizing pump in the pressurized state; (2) they are typically heavy and expensive due the use of materials such as stainless steel; and (3) it is difficult to regulate pressurization rate with hand pumps. Therefore, we developed a transportable high-pressure system suitable for cell preservation under high-pressure conditions. This high-pressure vessel has the following characteristics: (1) it can be easily separated from the pressurizing pump due to the use of a cock-type stop valve; (2) it is small and compact, is made of PEEK and weighs less than 200 g; and (3) pressurization rate is regulated by an electric pump instead of a hand pump. Using this transportable high-pressure vessel for cell preservation, we found that astrocytes can survive for 4 days at 1.6 MPa and 4 0 C.

78 FR 41369 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2013-07-10

... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Preliminary Results of..., line and pressure pipe (small diameter seamless pipe) from Romania. The period of review (POR) is... and Alloy Seamless Standard, Line and Pressure Pipe from Romania,'' dated concurrently with this...

Burst pressure investigation of filament wound type IV composite pressure vessel

Science.gov (United States)

Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

2017-12-01

Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

Mager, T.R.; Dragunov, Y.G.; Leitz, C.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2016-05-15

The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

Firefighter's compressed air breathing system pressure vessel development program

Science.gov (United States)

Beck, E. J.

1974-01-01

The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7

International Nuclear Information System (INIS)

Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.; Bryan, R.H.

1976-08-01

The test of intermediate test vessel V-7 was a crack-initiation fracture test of a 152-mm-thick (6-in.), 990-mm-OD (39-in.) vessel of ASTM A533, grade B, class 1 steel plate with a sharp outside surface flaw 457 mm (18 in.) long and about 135 mm (5.3 in.) deep. The vessel was heated to 91 0 C (196 0 F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

Leak isolation self-repairing tape for a water storage vessel and piping against holes

International Nuclear Information System (INIS)

Nagaya, Kosuke; Sekiguchi, Takahiro; Chen, Zhichao; Murakami, Iwanori

2007-01-01

A new type taping for a water storage vessel or piping is presented, in which water leakage is isolated automatically by its self-repairing mechanism against holes. The self-repairing unit (sealant layer) is consisting of three pieces of net with polymer particles inside lattices. Polymer particles, which expand their volume with water, is used for having self-repairing forces. In ordinary tapes, water leaks along the boundary between the tape and the vessel. In order to retain the leak isolation force, this article first discusses a method for making the sealant tape, then develops a method for fixing the sealant to the vessel. The portion of water leakage can be checked on this tape, and the method of detecting the hole or crack portion of the vessel is also presented by using the tape. (author)

Stress analysis of pressure vessels

International Nuclear Information System (INIS)

Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.

1979-01-01

This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)

Apparatus for carrying out ultrasonic inspection of pressure vessels

International Nuclear Information System (INIS)

Dent, K.H.; Greenhalgh, F.G.

1975-01-01

An apparatus is described for moving an ultrasonic scanning mechanism over the interior surface of a pressure vessel and comprising a mast for supporting the scanning mechanism inside the vessel and a carriage for traversing the mast within the vessel, the mast being pivotably secured to the carriage so that when the ultrasonic scanning mechanism contacts the interior surface of the pressure vessel the mast is caused to pivot. (auth)

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.)

76 FR 60083 - Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Japan and Romania

Science.gov (United States)

2011-09-28

... Alloy Seamless Standard, Line, and Pressure Pipe From Japan and Romania Determinations On the basis of... pressure pipe from Japan and Romania would be likely to lead to continuation or recurrence of material... regarding small- diameter carbon and alloy seamless standard, line, and pressure pipe from Romania...

Tribology aspects of a pressure vessel closure subjected to pressure cycling

International Nuclear Information System (INIS)

George, A.F.; Williams, M.E.

1988-04-01

A repair method being considered for a steel pressure vessel is to cut away the faulty part leaving an unreinforced circular hole in the curved wall and cover it with a sealed plate placed inside. In order to investigate the structural properties of such a repair a large model vessel (6m by 2m) was tested under pressure (about 2.5 MPa) and pressure cycling. This cycling caused relative movements at the loaded interface between the lid and the vessel. A tribological examination of the rubbing surfaces was carried out. The tribological examination is described and a small supporting programme of laboratory scaling tests. It gives the results and attempts to interpret them with particular attention given to wear, fretting fatigue and scaling to plant conditions. (author)

Residual Stress Estimation and Fatigue Life Prediction of an Autofrettaged Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Song, Kyung Jin; Kim, Eun Kyum; Koh, Seung Kee [Kunsan Nat’l Univ., Kunsan (Korea, Republic of)

2017-09-15

Fatigue failure of an autofrettaged pressure vessel with a groove at the outside surface occurs owing to the fatigue crack initiation and propagation at the groove root. In order to predict the fatigue life of the autofrettaged pressure vessel, residual stresses in the autofrettaged pressure vessel were evaluated using the finite element method, and the fatigue properties of the pressure vessel steel were obtained from the fatigue tests. Fatigue life of a pressure vessel obtained through summation of the crack initiation and propagation lives was calculated to be 2,598 cycles for an 80% autofrettaged pressure vessel subjected to a pulsating internal pressure of 424 MPa.

46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 97.30-1 Section... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer...

46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 196.30-1... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the Chief Engineer...

Dismantling method for reactor pressure vessel and system therefor

International Nuclear Information System (INIS)

Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

1994-01-01

Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head

Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) Nickel-Hydrogen Battery Performance Under LEO Cycling Conditions

Science.gov (United States)

Miller, Thomas B.; Lewis, Harlan L.

2004-01-01

LEO life cycle testing of Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) nickel-hydrogen cell packs have been sponsored by the NASA Aerospace Flight Battery Program. The cell packs have cycled under both 35% and 60% depth-of- discharge and temperature conditions of -5 C and +lO C. The packs have been on test since as early as 1992 and have generated a substantial database. This report will provide insight into performance trends as a function of the specific cell configuration and manufacturer for eight separate nickel-hydrogen battery cell packs.

Analysis code for pressure in reactor containment vessel of ATR. CONPOL

International Nuclear Information System (INIS)

1997-08-01

For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

46 CFR 176.812 - Pressure vessels and boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a.... (b) Periodic inspection and testing requirements for boilers are contained in § 61.05 in subchapter F...

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

In-service ultrasonic inspection of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Prepechal, J.; Sulc, J.

1982-01-01

Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

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)

Finite Element Limit Pressures for Circumferential Through-Wall Cracks in the Interface between Elbow and Pipe

International Nuclear Information System (INIS)

Jang, Yoon-Young; Han, Tae-Song; Huh, Nam-Su; Jeong, Jae-Uk

2014-01-01

Among integrity assessment method based on a fracture mechanics concept for piping system, a limit load method is one of the important way to predict a maximum load carrying capacity in the materials with high ductility in the sense that it is used to either assess directly structural integrity of pipe based on fully plastic fracture mechanics or calculate elasticplastic fracture mechanics parameters based on reference stress concept. In nuclear power plants, piping system often involves elbows welded to straight pipe. Since welded regions are vulnerable to cracking, it is important to predict an accurate limit load for pipes with a crack in the interface between elbows and attached pipes. However, although extensive works have been made for developing limit analysis methods for cracked pipes, they were mainly for straight pipes. Recently, limit moment solutions for elbow that is attached to straight pipe with a circumferential through-wall crack(TWC) in the interface were proposed, whereas limit pressure for this geometry is not suggested yet. In this context, plastic limit pressures of circumferential TWCs between elbow and straight pipe were calculated in the present study considering geometric parameters such as an elbow curvature, a pipe size and a crack length. In the present study, the FE plastic limit analyses for circumferential TWC in the interface between elbow and pipe under internal pressure were conducted based on elastic perfectly plastic assumption. Based on the present FE results, it is found that plastic limit pressures of straight pipes with circumferential TWC are not appropriate for predicting plastic limit pressures of circumferential TWC in the interface between elbow and pipe for shorter crack length

Common-Pressure-Vessel Nickel-Hydrogen Battery Development

OpenAIRE

Otzinger, Burton; Wheeler, James

1991-01-01

The dual-cell, common-pressure vessel, nickel-hydrogen configuration has recently emerged as an option for small satellite nickel-hydrogen battery application. An important incentive is that the dual-cell, CPV configured battery presents a 30 percent reduction in volume and nearly 50 percent reduction in mounting footprint, when compared with an equivalent battery of individual pressure- vessel (IPV) cells. In addition energy density and cost benefits are significant. Eagle-Picher Industries ...

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Nuclear reactor pressure vessel-specific flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.

1992-01-01

Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses

PWR pressure vessel integrity during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.

1981-01-01

Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

Seals for sealing a pressure vessel such as a nuclear reactor vessel or the like

International Nuclear Information System (INIS)

Bruns, H.J.; Huelsermann, K.H.

1975-01-01

A description is given of seals for sealing a pressure vessel such as a nuclear reactor vessel, steam boiler vessel, or any other vessel which is desirably sealed against pressure of the type including a housing and a housing closure that present opposed vertical sealing surfaces which define the sides of a channel. The seals of the present invention comprise at least one sealing member disposed in the channel, having at least one stop face, a base portion and two shank portions extending from the base portion to form a groove-like recess. The shank portions are provided with sealing surfaces arranged to mate with the opposed vertical pressure vessel sealing surfaces. A shank-spreading wedge element also disposed in the channel has at least one stop face and is engaged in the groove-like recess with the sealing member and wedge element stop face adjacent to each other

Design optimization of a thin walled pressure vessel

International Nuclear Information System (INIS)

Sadiq, S.

2001-01-01

Design evaluation of a pressure vessel is not only to build confidence on its integrity but also to reduce structural weight and enhance the performance of the structure. Pressure vessel, e.g., a rocket motor not only has to withstand the high operating temperatures but it must also be able to survive the internal pressures and external aerodynamic forces and bending stresses during its operation in flight. A research program was devised to study the stresses, which are generated in a thin walled pressure vessel during actual operation and its simulation with cold testing technique, i.e., by means of hydrostatic testing employing electrical resistance strain gauges on the external surface of the cylinder. The objective of the research was to uphold the performance of the vessel by reducing its thickness from 6.09 to 5.5 mm (which of course reduces the safety factor margin from 1.8 to 1.5); thereby curtailing the overall structural weight and maintaining the efficiency of the vessel itself during its live operation. The techniques employed were hydrostatic testing, data acquisition system for obtaining data on strains from the electrical resistance strain gauges and later employing V on Mises yield criterion empirical relation to computer the stresses in hoop and longitudinal directions. (author)

Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

International Nuclear Information System (INIS)

Chou, Hsoung-Wei; Huang, Chin-Cheng

2016-01-01

Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

Computation of the effect of pipe plasticity on pressure-pulse propagation in a fluid system

International Nuclear Information System (INIS)

Youngdahl, C.K.; Kot, C.A.

1975-04-01

A simple computational model is developed for incorporating the effect of elastic-plastic deformation of piping on pressure-transient propagation in a fluid system. A computer program (PLWV) is described that incorporates this structural interaction model into a one-dimensional method-of-characteristics procedure for fluid-hammer analysis. Computed results are shown to be in good agreement with available experimental data. The most significant effect of plastic deformation is to limit the peak pressure of a pulse leaving a pipe to approximately the yield pressure of the pipe, if the pipe is sufficiently long. 7 references. (U.S.)

GOLD PRESSURE VESSEL SEAL

Science.gov (United States)

Smith, A.E.

1963-11-26

An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

Design of pressure vessels using shape optimization: An integrated approach

Energy Technology Data Exchange (ETDEWEB)

Carbonari, R.C., E-mail: ronny@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Munoz-Rojas, P.A., E-mail: pablo@joinville.udesc.br [Department of Mechanical Engineering, Universidade do Estado de Santa Catarina, Bom Retiro, Joinville, SC 89223-100 (Brazil); Andrade, E.Q., E-mail: edmundoq@petrobras.com.br [CENPES, PDP/Metodos Cientificos, Petrobras (Brazil); Paulino, G.H., E-mail: paulino@uiuc.edu [Newmark Laboratory, Department of Civil and Environmental Engineering, University of Illinois at Urbana-Champaign, 205 North Mathews Av., Urbana, IL 61801 (United States); Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 158 Mechanical Engineering Building, 1206 West Green Street, Urbana, IL 61801-2906 (United States); Nishimoto, K., E-mail: knishimo@usp.br [Department of Naval Architecture and Ocean Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Silva, E.C.N., E-mail: ecnsilva@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil)

2011-05-15

Previous papers related to the optimization of pressure vessels have considered the optimization of the nozzle independently from the dished end. This approach generates problems such as thickness variation from nozzle to dished end (coupling cylindrical region) and, as a consequence, it reduces the optimality of the final result which may also be influenced by the boundary conditions. Thus, this work discusses shape optimization of axisymmetric pressure vessels considering an integrated approach in which the entire pressure vessel model is used in conjunction with a multi-objective function that aims to minimize the von-Mises mechanical stress from nozzle to head. Representative examples are examined and solutions obtained for the entire vessel considering temperature and pressure loading. It is noteworthy that different shapes from the usual ones are obtained. Even though such different shapes may not be profitable considering present manufacturing processes, they may be competitive for future manufacturing technologies, and contribute to a better understanding of the actual influence of shape in the behavior of pressure vessels. - Highlights: > Shape optimization of entire pressure vessel considering an integrated approach. > By increasing the number of spline knots, the convergence stability is improved. > The null angle condition gives lower stress values resulting in a better design. > The cylinder stresses are very sensitive to the cylinder length. > The shape optimization of the entire vessel must be considered for cylinder length.

77 FR 67336 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2012-11-09

... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Final Results of Antidumping... alloy seamless standard, line and pressure pipe from Romania. The period of review is August 1, 2010..., line and pressure pipe from Romania. See Certain Small Diameter Carbon and Alloy Seamless Standard...

78 FR 63164 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2013-10-23

... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Final Results of Antidumping... carbon and alloy seamless standard, line and pressure pipe from Romania. For the final results we... pressure pipe from Romania.\\1\\ We invited interested parties to comment on the Preliminary Results. We...

Prestressed concrete pressure vessels for boiling water reactors

International Nuclear Information System (INIS)

Menon, S.

1979-12-01

Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

Eddy current testing of composite pressure vessels

Science.gov (United States)

Casperson, R.; Pohl, R.; Munzke, D.; Becker, B.; Pelkner, M.

2018-04-01

The use of composite pressure vessels instead of conventional vessels made of steel or aluminum grew strongly over the last decade. The reason for this trend is the tremendous weight saving in the case of composite vessels. However, the long-time behavior is not fully understood for filling and discharging cycles and creep strength and their influence on the CFRP coating (carbon fiber reinforced plastics) and the internal liner (steel, aluminum, or plastics). The CFRP ensures the pressure resistance while the inner liner is used as a container for liquid or gas. To overcome the missing knowledge of aging, BAM started an internal project to investigate degradation of these material systems. Therefore, applicable testing methods like eddy current testing are needed. Normally, high-frequency eddy current testing (HF-ET, f > 10 MHz) is deployed for CFRP due to its low conductivity of the fiber, which is in the order of 0.01 MS/s, and the capacitive coupling between the fibers. Nevertheless, in some cases conventional ET can be applied. We show a concise summary of studies on the application of conventional ET of composite pressure vessels.

Pressurized water reactor with reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1985-01-01

The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

Pressurized water reactor with reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1980-01-01

The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

Fluid-structure dynamics; Proceedings of the pressure vessels and piping conference, New Orleans, LA, June 23-26, 1985

International Nuclear Information System (INIS)

Ma, D.C.; Moody, F.J.

1985-01-01

Aspects of seismic analysis and testing of fluid-structure systems are discussed, taking into account an earthquake response analysis method for a liquid-structure-ground coupled system using the finite element method, a seismic analysis of liquid-filled tanks with an eccentric core barrel, a study of the seismic response of fluid-coupled coaxial cylinder exciting by horizontal and vertical loading, the seismic analysis of fluid-structure systems including perforated circular plates on the basis of a use of the finite element method, and the uplifting of earthquake-loaded liquid-filled tanks. Computational methods for coupled fluid-structure analysis are considered along with the fluid structure wave motion, flow induced vibration, and the design for unsteady fluid flow in vessel and piping systems. Attention is given to hydraulic transient analysis as a tool in setting system control, a new model on transient wave propagation in fluid-filled tubes, an investigation of injector instability in rocket engine manifold, and the fluid structure response of axial cracked cylinders

Fourier series analysis of a cylindrical pressure vessel subjected to axial end load and external pressure

International Nuclear Information System (INIS)

Brar, Gurinder Singh; Hari, Yogeshwar; Williams, Dennis K.

2013-01-01

This paper presents the comparison of a reliability technique that employs a Fourier series representation of random axisymmetric and asymmetric imperfections in a cylindrical pressure vessel subjected to an axial end load and external pressure, with evaluations prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 Rules. The ultimate goal of the reliability technique described herein is to predict the critical buckling load associated with the subject cylindrical pressure vessel. Initial geometric imperfections are shown to have a significant effect on the calculated load carrying capacity of the vessel. Fourier decomposition was employed to interpret imperfections as structural features that can be easily related to various other types of defined imperfections. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion, which are availed in the form of a double Fourier series. Fifty simulated shells generated by the Monte Carlo technique are employed in the final prediction of the critical buckling load. The representation of initial geometrical imperfections in the cylindrical pressure vessel requires the determination of respective Fourier coefficients. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries in combination with asymmetric imperfections as a function of position within the given cylindrical shell. The probability of the ultimate buckling stress exceeding a predefined threshold stress is also calculated. The method and results described herein are in stark contrast to the “knockdown factor” approach as applied to compressive stress evaluations currently utilized in industry. Further effort is needed to improve on the current design rules regarding column buckling of large diameter pressure vessels subjected to an axial end load and external pressure designed in accordance with ASME Boiler and

Stress analysis and evaluation of a rectangular pressure vessel

International Nuclear Information System (INIS)

Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

1992-10-01

This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

Science.gov (United States)

2010-10-01

... 46 Shipping 3 2010-10-01 2010-10-01 false Repairs of boiler and pressure vessels. 78.33-1 Section... OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-1 Repairs of boiler and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer shall...

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

Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

Directory of Open Access Journals (Sweden)

Hereil Pierre-Louis

2015-01-01

Full Text Available Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

USER SPECIFICATIONS FOR PRESSURE VESSELS AND TECHNICAL INTEGRITY

Directory of Open Access Journals (Sweden)

K.S. Johnston

2012-01-01

Full Text Available

ENGLISH ABSTRACT: Specifications translated from user requirements are prescribed in an attempt to capture and incorporate best practices with regards to the design, fabrication, testing, and operation of pressure vessels. The question as to whether these requirements affect the technical integrity of pressure vessels is often a subjective matter. This paper examines typical user requirement specifications against technical integrity of pressure vessels.
The paper draws on a survey of a convenience sample of practising engineers in a diversified petrochemical company. When compared with failures on selected pressure vessels recorded by Phillips and Warwick, the respondent feedback confirms the user specifications that have the highest impact on technical integrity.

AFRIKAANSE OPSOMMING: Gebruikersbehoeftes word saamgevat in spesifikasies wat lei tot goeie praktyk vir ontwerp, vervaarding, toetsing en bedryf van drukvate. Subjektiwiteit van die gebruikersbehoeftes mag soms die tegniese integriteit van ‘n drukvat beinvloed.
Die navorsing maak by wyse van monsterneming gebruik van die kennis van ingenieurs wat werk in ‘n gediversifiseerde petrochemiese bedryf. Die terugvoering bevestig dat bogenoemde spesifikasies inderdaad die grootste invloed het op tegniese integriteit.

46 CFR 154.650 - Cargo tank and process pressure vessel welding.

Science.gov (United States)

2010-10-01

... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo tank and process pressure vessel welding. 154.650... Equipment Construction § 154.650 Cargo tank and process pressure vessel welding. (a) Cargo tank and process pressure vessel welding must meet Subpart 54.05 and Part 57 of this chapter. (b) Welding consumables used...

Development of high pressure pipe scanners

International Nuclear Information System (INIS)

Kim, Jae H.; Lee, Jae C.; Moon, Soon S.; Eom, Heung S.; Choi, Yu R.

1998-12-01

This report describes an automatic ultrasonic scanning system for pressure pipe welds, which was developed in this project using recent advanced technologies on mobile robot and computer. The system consists of two modules: a robot scanner module which navigates and manipulates scanning devices, and a data acquisition module which generates ultrasonic signal and processes the data from the scanner. The robot has 4 magnetic wheels and 2 -axis manipulator on which ultrasonic transducer attached. The wheeled robot can navigate curved surface such as outer wall of circular pipes. Magnetic wheels were optimally designed through magnetic field analysis. Free surface sensing and line tracking control algorithm were developed and implemented, and the control devices and software can be used in practical inspection works. We expect our system can contribute to reduction of inspection time, performance enhancement, and effective management of inspection results

Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

International Nuclear Information System (INIS)

Lee, K.B.; Im, I.Y.; Lee, S.K.

1992-01-01

An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

Internal Friction of Pressure Vessel Steel Embrittlement

International Nuclear Information System (INIS)

Van Ouytsel, K.

2001-01-01

The contribution consists of an abstract of a PhD thesis. The thesis contains a literature study, a description of the construction details of a new inverted torsion pendulum. This device was designed to investigate pressure-vessel steels at high amplitudes (10 -4 to 10 -2 ) and over a wide temperature range (90-700K) at approximately 1 Hz in the irradiated condition. Results of measurements on a variety of reactor pressure vessel steels by means of the torsion penduli are reported and interpreted

Analysis and Design of Cryogenic Pressure Vessels for Automotive Hydrogen Storage

Science.gov (United States)

Espinosa-Loza, Francisco Javier

Cryogenic pressure vessels maximize hydrogen storage density by combining the high pressure (350-700 bar) typical of today's composite pressure vessels with the cryogenic temperature (as low as 25 K) typical of low pressure liquid hydrogen vessels. Cryogenic pressure vessels comprise a high-pressure inner vessel made of carbon fiber-coated metal (similar to those used for storage of compressed gas), a vacuum space filled with numerous sheets of highly reflective metalized plastic (for high performance thermal insulation), and a metallic outer jacket. High density of hydrogen storage is key to practical hydrogen-fueled transportation by enabling (1) long-range (500+ km) transportation with high capacity vessels that fit within available spaces in the vehicle, and (2) reduced cost per kilogram of hydrogen stored through reduced need for expensive structural material (carbon fiber composite) necessary to make the vessel. Low temperature of storage also leads to reduced expansion energy (by an order of magnitude or more vs. ambient temperature compressed gas storage), potentially providing important safety advantages. All this is accomplished while simultaneously avoiding fuel venting typical of cryogenic vessels for all practical use scenarios. This dissertation describes the work necessary for developing and demonstrating successive generations of cryogenic pressure vessels demonstrated at Lawrence Livermore National Laboratory. The work included (1) conceptual design, (2) detailed system design (3) structural analysis of cryogenic pressure vessels, (4) thermal analysis of heat transfer through cryogenic supports and vacuum multilayer insulation, and (5) experimental demonstration. Aside from succeeding in demonstrating a hydrogen storage approach that has established all the world records for hydrogen storage on vehicles (longest driving range, maximum hydrogen storage density, and maximum containment of cryogenic hydrogen without venting), the work also

Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

International Nuclear Information System (INIS)

Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

2000-01-01

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

Pressure wave propagation in sodium loop

International Nuclear Information System (INIS)

Botelho, D.A.

1989-01-01

A study was done on the pressure wave propagation within the pipes and mixture vessel of a termohydraulic loop for thermal shock with sodium. It was used the characteristic method to solve the one-dimensional continuity and momentum equations. The numerical model includes the pipes and the effects of valves and other accidents on pressure losses. The study was based on designer informations and engineering tables. It was evaluated the pressure wave sizes, parametrically as a function of the draining valve closure times. (author) [pt

Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

International Nuclear Information System (INIS)

Cunha, Sérgio B.; Netto, Theodoro A.

2012-01-01

The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3″ diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: ► An analytical model for the burst of a pipe with a volumetric flaw is developed. ► Deformation, strain and stress are modeled in the elastic and plastic domains. ► The model is comprehensively validated by experiments and numerical simulations. ► The burst pressure model’s accuracy is equivalent to finite element simulations.

The need to pressure test prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Forgie, J.H.; Holland, J.A.

1983-01-01

In the period when PCRV were relatively unproven, proof pressure testing provided a useful demonstration of vessel integritiy and a confirmation of model testing and of analysis. No failures have occurred during concrete vessel tests in the UK or in the subsequent operational life of the vessels and much has been learned of their behaviour in service. The paper examines the advantages and disadvantages of proof testing PCRV in the light of the above increased knowledge of vessel performance. The paper draws attention to certain hypothetical loading cases that could be more onerous than the proof test and suggests that pressure testing could itself cause unnecessarily high loading to parts of the vessel. Always recognising the safety considerations and demonstrations of such are of prime importance, the authors suggest that a lower pressure level could be adopted without loss of original intent. In addition some ground rules are suggested as to cases where proof testing could be omitted. (orig./HP)

46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Report of repairs to boilers and pressure vessels. 109... Report of repairs to boilers and pressure vessels. Before making repairs, except normal repairs and maintenance such as replacement of valves or pressure seals, to boilers or unfired pressure vessels in...

Failure rates in piping manufactured to different standards

International Nuclear Information System (INIS)

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

1995-11-01

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

75 FR 69125 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

Science.gov (United States)

2010-11-10

... with material injury by reason of imports from China of certain seamless carbon and alloy steel standard, line, and pressure pipe (``seamless SLP pipe''), provided for in subheadings 7304.19.10, 7304.19... Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China Determination On the basis of...

46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as...

Integrity of Magnox reactor steel pressure vessels

International Nuclear Information System (INIS)

Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

1992-01-01

The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

Prediction of pressure between packers of staged fracturing pipe strings in high-pressure deep wells and its application

Directory of Open Access Journals (Sweden)

Fuxiang Zhang

2015-03-01

Full Text Available Addressing to the deteriorated load conditions of working string and packers caused by annular pressure drop between packers during the staged stimulation of high-pressure deep well, one 2D temperature field transient prediction model for borehole under injecting conditions which considers such influences as friction heat, convection heat exchange was set up, based on energy conservation principle and borehole heat transfer theory. By means of analyzing the influences of borehole temperature and pressure changes on the annular volume between packers, and in combination with borehole temperature transient prediction model, annular fluid PVT equations of state, radial deformation model of tubing and formation transient seepage equation, a typical high-pressure deep well inter-packer annular pressure prediction model was established. Taking a high-pressure gas well in Tarim Oilfield for example, the inter-packer annular pressure prediction was conducted, on which, the mechanical analysis on packers and working strings was carried out. The analysis results show that although the pipe string is safe in the viewpoint of conventional design methods, it is still susceptible to failure after the annular pressure drop between packers was taken into consideration. Such factor should be fully considered in the design of staged stimulation pipe strings, and this prediction model provides new thoughts for the optimal design of high-pressure deep well staged stimulation pipe strings.

Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

International Nuclear Information System (INIS)

Sun Yunxue; Shao Jin

2015-01-01

Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

Factors affecting the integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.

1983-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, if certain postulated accidents, referred to as overcooling accidents, were to occur, the pressure vessel could be subjected to severe thermal shock while the pressure is substantial. As a result, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner-surface flaws prior to the vessel's normal end of life. A fracture-mechanics analysis for a typical postulated accident and also related thermal-shock experiments indicate that very shallow surface flaws that extend through the cladding into the base material could propagate. This is of particular concern because shallow flaws appear to be the most probable and presumably are the most difficult to detect

78 FR 31574 - Welded Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam; Institution of...

Science.gov (United States)

2013-05-24

... INTERNATIONAL TRADE COMMISSION [Investigation Nos. 731-TA-1210-1212 (Preliminary)] Welded Stainless Steel Pressure Pipe From Malaysia, Thailand, and Vietnam; Institution of Antidumping Duty..., by reason of imports from Malaysia, Thailand, and Vietnam of welded stainless steel pressure pipe...

Code boiler and pressure vessel life assessment

International Nuclear Information System (INIS)

Farr, J.R.

1992-01-01

In the United States of America and in Canada, laws and controls for determining life assessment for continued operation of equipment exist only for those pressure vessels built to Section III and evaluated according to Section XI. In this presentation, some of those considerations which are made in the USA and Canada for deciding on life or condition assessment of boilers and pressure vessels designed and constructed to other sections of the ASME Boiler and Pressure Vessel Code are reviewed. Life assessment or condition assesssment is essential in determining what is necessary for continued operation. With no ASME rules being adopted by laws or regulations, other than OSHA in the USA and similar environmental controls in Canada, to control life assessment for continued operation, the equipment owner must decide if assessment is to be done and how much to do. Some of those considerations are reviewed along with methods and procedures to make an assessment along with a discussion of where the ASME B and PV Code currently stands regarding continued operation. (orig.)

Increase of cyclic durability of pressure vessels

International Nuclear Information System (INIS)

Vorona, V.A.; Zvezdin, Yu.I.

1980-01-01

The durability of multilayer pressure vessels under cyclic loading is compared with single-layer vessels. The relative conditional durability is calculated taking into account the assumption on the consequent destruction of layers and viewing a vessel wall as an indefinite plate. It is established that the durability is mainly determined by the number of layers and to a lesser degree depends on the relative size of the defect for the given layer thickness. The advantage of the multilayer vessels is the possibility of selecting layer materials so that to exclude the effect of agressive corrosion media on the strength [ru

Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

Energy Technology Data Exchange (ETDEWEB)

Cunha, Sergio B., E-mail: sbcunha@petrobras.com.br [PETROBRAS/TRANSPETRO, Av. Pres. Vargas 328 - 7th floor, Rio de Janeiro, RJ 20091-060 (Brazil); Netto, Theodoro A., E-mail: tanetto@lts.coppe.ufrj.br [COPPE, Federal University ot Rio de Janeiro, Ocean Engineering Department, PO BOX 68508, Rio de Janeiro - RJ (Brazil)

2012-01-15

The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3 Double-Prime diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: Black-Right-Pointing-Pointer An analytical model for the burst of a pipe with a volumetric flaw is developed. Black-Right-Pointing-Pointer Deformation, strain and stress are modeled in the elastic and plastic domains. Black-Right-Pointing-Pointer The model is comprehensively validated by experiments and numerical simulations. Black-Right-Pointing-Pointer The burst pressure model's accuracy is equivalent to finite element simulations.

Pressure vessel and method therefor

Science.gov (United States)

Saunders, Timothy

2017-09-05

A pressure vessel includes a pump having a passage that extends between an inlet and an outlet. A duct at the pump outlet includes at least one dimension that is adjustable to facilitate forming a dynamic seal that limits backflow of gas through the passage.

Computer program TMOC for calculating of pressure transients in fluid filled piping networks

International Nuclear Information System (INIS)

Siikonen, T.

1978-01-01

The propagation of a pressure wave in fluid filles tubes is significantly affected by the pipe wall motion and vice versa. A computer code TMOC (Transients by the Method of Characteristics) is being developed for the analysis of the coupled fluid and pipe wall transients. Because of the structural feedback, the pressure can be calculated more accurately than in the programs commonly used. (author)

Pressure-dependent fragilities for piping components: Pilot study on Davis-Besse Nuclear Power Station

International Nuclear Information System (INIS)

Wesley, D.A.; Nakaki, D.K.; Hadidi-Tamjed, H.; Kipp, T.R.

1990-10-01

The capacities of four, low-pressure fluid systems to withstand pressures and temperatures above the design levels were established for the Davis-Besse Nuclear Power Station. The results will be used in evaluating the probability of plant damage from Interfacing System Loss of Coolant Accidents (ISLOCA) as part of the probabilistic risk assessment of the Davis-Besse nuclear power station undertaken by EG ampersand G Idaho, Inc. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The probabilities of failure, as a function of internal pressure, are evaluated as well as the variabilities associated with them. Leak rates or leak areas are estimated for the controlling modes of failure. The pressure capacities for the pipes and vessels are evaluated using limit-state analyses for the various failure modes considered. The capacities are dependent on several factors, including the material properties, modeling assumptions, and the postulated failure criteria. The failure modes for gasketed-flange connections, valves, and pumps do not lend themselves to evaluation by conventional structural mechanics techniques and evaluation must rely primarily on the results from ongoing gasket research test programs and available vendor information and test data. 21 refs., 7 figs., 52 tabs

Investigation of two-phase flow structure in model of draught pipe of water boiling reactor VK-300

International Nuclear Information System (INIS)

Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P.

2001-01-01

VK-300 reactor represents a vessel-type boiling reactor with integral arrangement of assemblies and in-vessel steam separation at one-circuit scheme. The circuit consists of core, draught pipes, and separation facilities. The vessel of VK-300 reactor is chosen on the base of the dimensions of that of VVER-1000 reactor. The following thermal-hydraulic parameters of nuclear power plant (NPP) were investigated experimentally: dependence of void fraction upon the steam quality in mixing chamber (on the draught section input); pressure losses at different, specific zones of up-flow and down-flow sections of the circuit with free circulation; degree of steam separation in the separating chamber (at the first step of phase separation) and its dependence upon steam quality; structure of steam-water flow in draught pipes (distribution of phases over the draught pipe cross- section); presence of steam hovering and height of this hovering in inter-pipe space of draught section. (author)

Heavy-Section Steel Technology Program intermediate-scale pressure vessel tests

International Nuclear Information System (INIS)

Bryan, R.H.; Merkle, J.G.; Smith, G.C.; Whitman, G.D.

1977-01-01

The tests of intermediate-size vessels with sharp flaws permitted the comparison of experimentally observed behavior with analytical predictions of the behavior of flawed pressure vessels. Fracture strains estimated by linear elastic fracture mechanics (LEFM) were accurate in the cases in which the flaws resided in regions of high transverse restraint and the fracture toughness was sufficiently low for unstable fracture to occur prior to yielding through the vessel wall. When both of these conditions were not present, unstable fracture did occur, always preceded by stable crack growth; and the cylinders with flaws initially less than halfway through the wall attained gross yield prior to burst. Predictions of failure pressure of the vessels with flawed nozzles, based upon LEFM estimates of failure strain, were very conservative. LEFM calculations of critical load were based upon small-specimen fracture toughness test data. Whenever gross yielding preceded failure, the actual strains achieved were considerably greater than the estimated strains at failure based on LEFM. In such cases the strength of the vessel may be no longer dependent upon plane-strain fracture toughness but upon the capacity of the cracked section to carry the imposed load stably in the plastic range. Stable crack growth, which has not been predictable quantitatively, is an important factor in elastic-plastic analysis of strength. The ability of the flawed vessels to attain gross yield in unflawed sections has important qualitative implications on pressure vessel safety margins. The gross yield condition occurs in light-water-reactor pressure vessels at about 2 x design pressure. The intermediate vessel tests that demonstrated a capacity for exceeding this load confirm that the presumed margin of safety is not diminished by the presence of flaws of substantial size, provided that material properties are adequate

Rapid construction of concrete pressure vessels

International Nuclear Information System (INIS)

Limbert, D.; Weatherseed, D.C.

1989-01-01

This paper opens with a general description of the concrete pressure vessel followed by a more detailed examination of the critical elements of the construction, including choice of methods and plant which were selected to ensure its rapid construction. The pressure vessel construction cannot be treated in isolation, because it is very closely linked with its surrounding structures - namely the reactor hall which surrounds it and the charge hall which tops it, as will be seen in the context of this paper. Rate of progress of construction is not entirely in the civil contractor's hands because so many of the operations affecting the civil works are of a mechanical nature, hence a very close liaison and understanding amongst all contractors concerned was of the utmost importance. (author)

Light Water Reactor-Pressure Vessel Surveillance project computer system

International Nuclear Information System (INIS)

Merriman, S.H.

1980-10-01

A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

PPOOLEX experiments with a modified blowdown pipe outlet

International Nuclear Information System (INIS)

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

2009-08-01

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

PPOOLEX experiments with a modified blowdown pipe outlet

Energy Technology Data Exchange (ETDEWEB)

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

2009-08-15

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

78 FR 62583 - Welded Stainless Pressure Pipe From Malaysia, Thailand, and the Socialist Republic of Vietnam...

Science.gov (United States)

2013-10-22

... Pressure Pipe From Malaysia: Request for Extension of Preliminary Determination,'' ``Welded Stainless Steel... Stainless Pressure Pipe From Malaysia, Thailand, and the Socialist Republic of Vietnam: Postponement of...: Charles Riggle (Malaysia), Brandon [[Page 62584

Survey on application of probabilistic fracture mechanics approach to nuclear piping

International Nuclear Information System (INIS)

Kashima, Koichi

1987-01-01

Probabilistic fracture mechanics (PFM) approach is newly developed as one of the tools to evaluate the structural integrity of nuclear components. This report describes the current status of PFM studies for pressure vessel and piping system in light water reactors and focuses on the investigations of the piping failure probability which have been undertaken by USNRC. USNRC reevaluates the double-ended guillotine break (DEGB) of rector coolant piping as a design basis event for nuclear power plant by using the PFM approach. For PWR piping systems designed by Westinghouse, two causes of pipe break are considered: pipe failure due to the crack growth and pipe failure indirectly caused by failure of component supports due to an earthquake. PFM approach shows that the probability of DEGB from either cause is very low and that the effect of earthquake on pipe failure can be neglected. (author)

A novel numerical model for estimating the collapse pressure of flexible pipes

Energy Technology Data Exchange (ETDEWEB)

Nogueira, Victor P.P.; Antoun Netto, Theodoro [Universidade Federal do Rio de Janeiro (COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao em Engenharia], e-mail: victor@lts.coppe.ufrj.br

2009-07-01

As the worldwide oil and gas industry operational environments move to ultra-deep waters, failure mechanisms in flexible pipes such as instability of the armor layers under compression and hydrostatic collapse are more likely to occur. Therefore, it is important to develop reliable numerical tools to reproduce the failure mechanisms that may occur in flexible pipes. This work presents a representative finite element model of flexible pipe capable to reproduce its pre and post-collapse behavior under hydrostatic pressure. The model, developed in the scope of this work, uses beam elements and includes nonlinear kinematics and material behavior influences. The dependability of the numerical results is assessed in light of experimental tests on flexible pipes with 4 inches and 8 inches nominal diameter available in the literature (Souza, 2002). The applied methodology provided coherent values regarding the estimation of the collapse pressures and results have shown that the proposed model is capable to reproduce experimental results. (author)

76 FR 47555 - Certain Large Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Japan...

Science.gov (United States)

2011-08-05

... Engineers (``ASME'') code stress levels. Alloy pipes made to ASTM A-335 standard must be used if temperatures and stress levels exceed those allowed for ASTM A-106. Seamless pressure pipes sold in the United..., or pressure pipe applications. B. Finished and unfinished oil country tubular goods (``OCTG''), if...

Multiple cell common pressure vessel nickel hydrogen battery

Science.gov (United States)

Zagrodnik, Jeffrey P.; Jones, Kenneth R.

1991-01-01

A multiple cell common pressure vessel (CPV) nickel hydrogen battery was developed that offers significant weight, volume, cost, and interfacing advantages over the conventional individual pressure vessel (IPV) nickel hydrogen configuration that is currently used for aerospace applications. The baseline CPV design was successfully demonstrated though the testing of a 26 cell prototype, which completed over 7,000 44 percent depth of discharge LEO cycles. Two-cell boilerplate batteries have now exceeded 12,500 LEO cycles in ongoing laboratory tests. CPV batteries using both nominal 5 and 10 inch diameter vessels are currently available. The flexibility of the design allows these diameters to provide a broad capability for a variety of space applications.

Sludge pipe flow pressure drop prediction using composite power ...

African Journals Online (AJOL)

2011-09-30

Sep 30, 2011 ... 3Department of Chemical Engineering, IIT Kanpur, India. Abstract. When predicting pressure gradients for the flow of sludges in pipes, the rheology of the fluid ..... implicit in the stability analysis of Ryan and Johnson (1959).

Pressure vessel lid

International Nuclear Information System (INIS)

Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

1986-01-01

The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

Experimental verification of integrated pressure suppression systems in fusion reactors at in-vessel loss-of-coolant events

International Nuclear Information System (INIS)

Takase, K.; Akimoto, H.

2001-01-01

An integrated ICE (Ingress-of-Coolant Event) test facility was constructed to demonstrate that the ITER safety design approach and design parameters for the ICE events are adequate. Major objectives of the integrated ICE test facility are: to estimate the performance of an integrated pressure suppression system; to obtain the validation data for safety analysis codes; and to clarify the effects of two-phase pressure drop at a divertor and the direct-contact condensation in a suppression tank. A scaling factor between the test facility and ITER-FEAT is around 1/1600. The integrated ICE test facility simulates the ITER pressure suppression system and mainly consists of a plasma chamber, vacuum vessel, simulated divertor, relief pipe and suppression tank. From the experimental results it was found quantitatively that the ITER pressure suppression system is very effective to reduce the pressurization due to the ICE event. Furthermore, it was confirmed that the analytical results of the TRAC-PF1 code can simulate the experimental results with high accuracy. (author)

Finite element limit analysis based plastic limit pressure solutions for cracked pipes

International Nuclear Information System (INIS)

Shim, Do Jun; Huh, Nam Su; Kim, Yun Jae; Kim, Young Jin

2002-01-01

Based on detailed FE limit analyses, the present paper provides tractable approximations for plastic limit pressure solutions for axial through-wall cracked pipe; axial (inner) surface cracked pipe; circumferential through-wall cracked pipe; and circumferential (inner) surface cracked pipe. Comparisons with existing analytical and empirical solutions show a large discrepancy in circumferential short through-wall cracks and in surface cracks (both axial and circumferential). Being based on detailed 3-D FE limit analysis, the present solutions are believed to be the most accurate, and thus to be valuable information not only for plastic collapse analysis of pressurised piping but also for estimating non-linear fracture mechanics parameters based on the reference stress approach

Manipulator for testing a top-opened reactor pressure vessel

International Nuclear Information System (INIS)

Bauer, R.; Kastl, H.

1991-01-01

The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

Reactor Pressure Vessel (RPV) Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

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.

Piping system damping data at higher frequencies

International Nuclear Information System (INIS)

Ware, A.G.

1987-01-01

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

Reactor pressure vessel steels

International Nuclear Information System (INIS)

Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

1998-01-01

Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

limit loads for wall-thinning feeder pipes under combined bending and internal pressure

International Nuclear Information System (INIS)

Je, Jin Ho; Lee, Kuk Hee; Chung, Ha Joo; Kim, Ju Hee; Han, Jae Jun; Kim, Yun Jae

2009-01-01

Flow Accelerated Corrosion (FAC) during inservice conditions produces local wall-thinning in the feeder pipes of CANDU. The Wall-thinning in the feeder pipes is main degradation mechanisms affecting the integrity of piping systems. This paper discusses the integrity assessment of wall-thinned feeder pipes using limit load analysis. Based on finite element limit analyses, this paper compare limit loads for wall-thinning feeder pipes under combined bending and internal pressure with proposed limit loads. The limit loads are determined from limit analyses based on rectangular wall-thinning and elastic-perfectly-plastic materials using the large geometry change.

30 CFR 57.13001 - General requirements for boilers and pressure vessels.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

30 CFR 56.13001 - General requirements for boilers and pressure vessels.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with the...

Pressurized-thermal-shock experiments with thick vessels

International Nuclear Information System (INIS)

Bryan, R.H.; Nanstad, R.K.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.

1986-01-01

Information is provided on the series of pressurized-thermal-shock experiments at the Oak Ridge National Laboratory, motivated by a concern for the behavior of flaws in reactor pressure vessels having welds or shells exhibiting low upper-shelf Charpy impact energies, approx. 68J or less

Safety of steel vessel Magnox pressure circuits

International Nuclear Information System (INIS)

Stokoe, T.Y.; Bolton, C.J.; Heffer, P.J.H.

1991-01-01

The maintenance of pressure circuit integrity is fundamental to nuclear safety at the steel vessel Magnox stations. To confirm continued pressure circuit integrity the CEGB, as part of the Long Term Safety Review, has carried out extensive assessment and inspection in recent years. The assessment methods and inspection techniques employed are based on the most modern available. Reactor pressure vessel integrity is confirmed by a combination of arguments including safety factors inferred from the successful pre-service overpressure test, leak-before-break analysis and probabilistic assessment. In the case of other parts of the pressure circuits that are more accessible, comprising the boiler shells and interconnecting gas duct work, in-service inspection is a major element of the safety substantiation. The assessment and inspection techniques and the materials property data have been underpinned for many years by extensive research and development programmes and in-reactor monitoring of representative samples has also been undertaken. The paper summarises the work carried out to demonstrate the long term integrity of the Magnox pressure circuits and provides examples of the results obtained. (author)

Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading

International Nuclear Information System (INIS)

Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

1978-01-01

HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7

Variability of mechanical properties of nuclear pressure vessel steels

International Nuclear Information System (INIS)

Petrequin, P.; Soulat, P.

1980-01-01

Causes of variability of mechanical properties nuclear pressure vessel steels are reviewed and discussed. The effects of product shape and size, processing history and heat treatment are investigated. Some quantitative informations are given on the scatter of mechanical properties of typical pressure vessel components. The necessity of using recommended or standardized properties for comparing mechanical properties before and after irradiation in pin pointed. (orig.) [de

Useful effectiveness of plastic pipes for gas pipelines operating at low pressure

Energy Technology Data Exchange (ETDEWEB)

Zayutsev, K I

1976-06-01

Because a significant portion of the feeder lines in the Soviet union operate at relatively low pressures up to 75 to 150 psi, it is economically feasible to replace the conventional pipeline material (steel) with plastic. Cost savings result from lower material costs and ease of laying plastic pipe. Because of stiffness and corrosion requirements, the steel pipe used for these low-pressure pipelines is much thicker than needed to withstand the pressure used. Data are tabulated on the comparative costs and manpower requirements for the construction of 1 km of steel, polyvinyl chloride, and polyethylene gas pipelines ranging in diameter from 3 to 14 in. Generally, the plastic pipelines required 15 to 30% less man-days and were 20 to 35% lower in cost to build. The plastic pipelines can operate at 150 to 175 psi pressure and at temperatures up to 100 to 140/sup 0/F. In research conducted at VNIIST (All-Union Research Institute for the Construction of Trunk Pipelines) on plastic pipelines, a 2.5-mile test section has been operating successfully for 2 years, and new techniques and equipment for joining plastic pipe up to 25-in. diameter are being developed.

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

FLANGE-ORNL, Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature

International Nuclear Information System (INIS)

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

1979-01-01

1 - Description of problem or function: FLANGE-ORNL calculates appropriate loads, stresses, and displacements for the flanges, bolts, and gaskets that comprise a flanged piping joint for internal pressure or moment loading on the pipe, temperature difference between the flange hub and ring, and variations in bolt load that result from pressure, hub-ring temperature gradient and/or bolt-ring temperature differences. Flanges considered may be tapered-hub, straight or blind. 2 - Method of solution: The solution is based on discontinuity analysis and the theory of plates and shells

Weld evaluation on spherical pressure vessels using holographic interferometry

International Nuclear Information System (INIS)

Boyd, D.M.; Wilcox, W.W.

1980-01-01

Waist welds on spherical experimental pressure vessels have been evaluated under pressure using holographic interferometry. A coincident viewing and illumination optical configuration coupled with a parabolic mirror was used so that the entire weld region could be examined with a single hologram. Positioning the pressure vessel at the focal point of the parabolic mirror provides a relatively undistorted 360 degree view of the waist weld. Double exposure and real time holography were used to obtain displacement information on the weld region. Results are compared with radiographic and ultrasonic inspections

The relevance of crack arrest phenomena for pressure vessel structural integrity assessment

International Nuclear Information System (INIS)

Connors, D.C.; Dowling, A.R.; Flewitt, P.E.J.

1996-01-01

The potential role of a crack arrest argument for the structural integrity assessments of steel pressure vessels and the relationship between crack initiation and crack arrest philosophies are described. A typical structural integrity assessment using crack initiation fracture mechanics is illustrated by means of a case study based on assessment of the steel pressure vessels for Magnox power stations. Evidence of the occurrence of crack arrest in structures is presented and reviewed, and the applications to pressure vessels which are subjected to similar conditions are considered. An outline is given of the material characterisation that would be required to undertake a crack arrest integrity assessment. It is concluded that crack arrest arguments could be significant in the structural integrity assessment of PWR reactor pressure vessels under thermal shock conditions, whereas for Magnox steel pressure vessels it would be limited in its potential to supporting existing arguments. (author)

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

International Nuclear Information System (INIS)

Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

1983-07-01

The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

Analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A.N. de; Augusto, O.B.

1985-01-01

This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

Design, fabrication and quality assurance of pressure vessels

International Nuclear Information System (INIS)

Kimura, Ichiro; Miki, Masao; Yamazaki, Tsuneji; Tanaka, Yoshikazu; Sato, Misao

1978-01-01

The production facilities, design and manufacturing technologies, and quality assurance in the Toyo Works, Ehime Manufactory, Sumitomo Heavy Industries, Ltd., which manufactures pressure vessels, are described, and especially the actual example of non-destructive tests is shown. The Toyo Works was completed in April, 1973, to manufacture large structures such as pressure vessels, offshore structures and bridges. The total area of the site is 535,000 m 2 , that of factory buildings is 33,600 m 2 , and the outdoor assembling yard is 114,800 m 2 . The large dry dock and main installations such as 12,000 tf hydraulic press, an annealing furnace, a heat treating furnace, a quenching tank, a horizontal boring machine, 6 m vertical lathe, various welding machines, 8 MeV X-ray apparatus, sand blasting and pickling facilities, and two 160 t cranes for shipment are arranged so as to enable smooth flow of production. The standards for chemical pressure vessels in various countries are compared, and considerably high allowable stress is adopted in Europe. The design and stress analysis of pressure vessels are carried out in accordance with ASME Section 8, Div. 1 or Div. 2. As for the materials, attention must be paid to the change of properties due to heat and strain, temper brittleness, low temperature toughness and so on. The quality assurance system must be established to observe the requirements of standards. (Kako, I.)

An experimental investigation on the pressure characteristics of high speed self-resonating pulsed waterjets influenced by feeding pipe diameter

Energy Technology Data Exchange (ETDEWEB)

Li, Dong; Kang, Dong; Ding, Xiao Long; Wang, Xiao Huan; Fang, Zhen Long [School of Power and Mechanical Engineering, Wuhan University, Hubei Province (China)

2016-11-15

The destructive power of a continuous waterjet issuing from a nozzle can be greatly enhanced by generating self-resonance in the nozzle assembly to produce a Self-resonating pulsed waterjet (SRPW). To further improve the performance of SRPW, effects of feeding pipe diameter on the pressure characteristics were experimentally investigated by measuring and analyzing the axial pressure oscillation peaks and amplitudes. Four organ-pipe nozzles of different chamber lengths and three feeding pipes of different diameters were employed. Results show that feeding pipe diameter cannot change the feature of SRPW of having an optimum standoff distance, but it slightly changes the oscillating frequency of the jet. It is also found that feeding pipe diameter significantly affects the magnitudes of pressure oscillation peak and amplitude, largely depending on the pump pressure and standoff distance. The enhancement or attenuation of the pressure oscillation peak and amplitude can be differently affected by the same feeding pipe diameter.

Contribution for the improvement of pressurized thermal shock assessment methodologies in PWR pressure vessels

International Nuclear Information System (INIS)

Gomes, Paulo de Tarso Vida

2005-01-01

The structural integrity assessment of nuclear reactor pressure vessel, concerned to Pressurized Thermal Shock (PTS) accidents, became a necessity and has been investigated since the eighty's. The recognition of the importance of PTS assessment has led the international nuclear technology community to devote a considerable research effort directed to the complete integrity assessment process of the Reactor Pressure Vessels (VPR). Researchers in Europe, Japan and U.S.A. have concentrated efforts in the VPR structural and fracture analysis, conducting experiments to best understand how specific factors act on the behavior of discontinuities, under PTS loading conditions. The main goal of this work is to study de structural behavior of an 'in scale' PWR nuclear reactor pressure vessel model, containing actual discontinuities, under loading conditions generated by a pressurized thermal shock. To construct the pressure vessel model utilized in this research, the approach developed by Barroso (1995) and based on likelihood studies, related to thermal-hydraulic behavior during the PTS was employed. To achieve the objective of this research, a new methodology to generate cracks, with known geometry and localization in the vessel model wall was developed. Additionally, an hydraulic circuit, able to flood the vessel model, heated to 300 deg C, with 10 m 3 of water at 8 deg C, in 170 seconds, was built. Thermo-hydraulic calculations using RELAP5/M0D 3.2.2γ computational code were done, to estimate the temperature profiles during the cooling time. The resulting data subsidized the thermo-structural calculations that were accomplished using ANSYS 7.01 computational code, for both 2D and 3D models. So, the stress profiles obtained with these calculations were associated with fracture mechanics concepts, to assess the crack growth behavior in the VPR model wall. After the PTS test, the VPR model was submitted to destructive and non-destructive inspections. The results

H.B. Robinson-2 pressure vessel benchmark

Energy Technology Data Exchange (ETDEWEB)

Remec, I.; Kam, F.B.K.

1998-02-01

The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

Welding in repair of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Pilous, V.; Kovarik, R.

1987-01-01

Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

Stress categorization in nozzle to pressure vessel connections finite elements models

International Nuclear Information System (INIS)

Albuquerque, Levi Barcelos de

1999-01-01

The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae for simple shell

Problems in Pressure Vessel Design and Manufacture

Energy Technology Data Exchange (ETDEWEB)

Hellstroem, O [Uddeholms AB, Degerfors (Sweden); Nilson, Ragnar [AB Atomenergi, Nykoeping (Sweden)

1963-05-15

The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels.

Problems in Pressure Vessel Design and Manufacture

International Nuclear Information System (INIS)

Hellstroem, O.; Nilson, Ragnar

1963-05-01

The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels

Development of computational methods of design by analysis for pressure vessel components

International Nuclear Information System (INIS)

Bao Shiyi; Zhou Yu; He Shuyan; Wu Honglin

2005-01-01

Stress classification is not only one of key steps when pressure vessel component is designed by analysis, but also a difficulty which puzzles engineers and designers at all times. At present, for calculating and categorizing the stress field of pressure vessel components, there are several computation methods of design by analysis such as Stress Equivalent Linearization, Two-Step Approach, Primary Structure method, Elastic Compensation method, GLOSS R-Node method and so on, that are developed and applied. Moreover, ASME code also gives an inelastic method of design by analysis for limiting gross plastic deformation only. When pressure vessel components design by analysis, sometimes there are huge differences between the calculating results for using different calculating and analysis methods mentioned above. As consequence, this is the main reason that affects wide application of design by analysis approach. Recently, a new approach, presented in the new proposal of a European Standard, CEN's unfired pressure vessel standard EN 13445-3, tries to avoid problems of stress classification by analyzing pressure vessel structure's various failure mechanisms directly based on elastic-plastic theory. In this paper, some stress classification methods mentioned above, are described briefly. And the computational methods cited in the European pressure vessel standard, such as Deviatoric Map, and nonlinear analysis methods (plastic analysis and limit analysis), are depicted compendiously. Furthermore, the characteristics of computational methods of design by analysis are summarized for selecting the proper computational method when design pressure vessel component by analysis. (authors)

Further fields of application for prestressed cast iron pressure vessels (PCIV)

International Nuclear Information System (INIS)

Guelicher, L.; Schilling, F.E.

1977-01-01

The redundancy of the prestressing system of prestressed structures as well as the clear separation of sealing and load-carrying functions of prestressed cast iron pressure vessels offer substantial advantages over conventional welded steel pressure vessels. Because of the temperature resistance of cast iron up to 400 0 C it is possible to build prestressed pressure vessels commercially as hot-working structures. The compressive strength of cast iron, which is 25 times as high as that of concrete allows for a very compact design of the PCIV. Further specific properties of the PCIV like pre-fabrication of the vessel in the production plant - made possible by a structure assembled from segments - short assembly periods at the construction site etc., may open more fields of application. - PCIV as pressurized storage tanks for the emergency shut down system in nuclear power stations. - PCIV as high pressure vessel for the chemical industry. - PCIV as energy storage. - PCIV for light water reactors. - PCIV as burst protection. It is concluded that the application of prestressed cast iron promises to be successful where either structures with large volumes and high pressures and/or temperatures are required or where aspects of safety allow for efficient use of prestressed structures. (Auth.)

Interpretation of strain measurements on nuclear pressure vessels

International Nuclear Information System (INIS)

Andersen, S.I.; Engbaek, P.

1979-11-01

Selected results from strain measurements on 4 nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzels, internal vessel structure and flange bolts. The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions. (author)

Renovation of the sealing planes of WWER-400 reactors pressure vessel

International Nuclear Information System (INIS)

Jablonicky, P.; Pilat, P.

2007-01-01

An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

Structural analysis and evaluation for the design of pressure vessel

International Nuclear Information System (INIS)

Arai, K.; Uragami, K.; Funada, T.; Baba, K.; Kira, T.

1977-01-01

For the design of pressure vessel, the detailed structural analysis such as the fatigue analysis under operating conditions is required by ASME Code or Japanese regulation. Accordingly, it should be verified by the analysis that the design of the pressure vessel is in compliance with the stress limitation defined in the Code or the regulation. However, it was apparent that the analysis is very complicated and takes a lot of time to evaluate in accordance with the Code requirements. Thereupon we developed the computer program by which we can perform the stress analysis with correctness and comparatively in a short period of design work reflecting the calculation results on detailed drawings to be used for fabrication. The computer program is controlled in combination with the system of the design work and out put list of the program can be directly used for the stress analysis report which is issued to customers. In addition to the above computer program, we developed the specific three dimensional finite element computer program to make sure of the structural integrity of the vessel head and flanges which are most complex for the analysis compared with the stress distribution measured by strain gauges on the vessel head and flange. Besides the structural analysis, the fracture mechanics analysis for the purpose of preventing the pressure vessel from the brittle fracture during heat-up and cool-down operation is also important and thereby we showed herein that the pressure vessel is in safety against the brittle fracture for the specified operating conditions. As a result of the above-mentioned analysis, the pressure vessel is designed with safety from the stand-points of the structural intensity and the fracture mechanics. (auth.)

In-service supervision of a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Zemann, H.; Mayer, N.; Amberg, C.

1985-01-01

On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C/50 bar). (Author)

In-service supervision of a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Zemann, H.; Weissbacher, L.; Mayer, N.; Amberge, C.

1985-01-01

On-line measurements of the physical state of a prestressed concrete pressure vessel, and comparison with the design predictions of the distribution of temperature, strain and stress within the concrete member and the criteria of layout, provide an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed with reference to the prototype vessel at Seibersdorf Research Centre during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C, 50 bar). (author)

Bounding the conservatism in flaw-related variables for pressure vessel integrity analyses

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.

1993-01-01

The fracture mechanics-based integrity analysis of a pressure vessel, whether performed deterministically or probabilistically, requires use of one or more flaw-related input variables, such as flaw size, number of flaws, flaw location, and flaw type. The specific values of these variables are generally selected with the intent to ensure conservative predictions of vessel integrity. These selected values, however, are largely independent of vessel-specific inspection results, or are, at best, deduced by ''conservative'' interpretation of vessel-specific inspection results without adequate consideration of the pertinent inspection system performance (reliability). In either case, the conservatism associated with the flaw-related variables chosen for analysis remains examination (NDE) technology and the recently formulated ASME Code procedures for qualifying NDE system capability and performance (as applied to selected nuclear power plant components) now provides a systematic means of bounding the conservatism in flaw-related input variables for pressure vessel integrity analyses. This is essentially achieved by establishing probabilistic (risk)-based limits on the assigned variable values, dependent upon the vessel inspection results and on the inspection system unreliability. Described herein is this probabilistic method and its potential application to: (i) defining a vessel-specific ''reference'' flaw for calculating pressure-temperature limit curves in the deterministic evaluation of pressurized water reactor (PWR) reactor vessels, and (ii) limiting the flaw distribution input to a PWR reactor vessel-specific, probabilistic integrity analysis for pressurized thermal shock loads

Periodic inspection for safety of CANDU heat transport piping systems

International Nuclear Information System (INIS)

Ellyin, F.

1979-10-01

Periodic inspection of heat transport and emergency core cooling piping systems is intended to maintain an adequate level of safety throughout the life of the plant, and to protect plant personnel and the public from the consequences of a failure and release of fission products. This report outlines a rational approach to the periodic inspection based on a fully probabilistic model. It demonstrates the methodology based on theoretical treatment and experimental data whereby the strength of a pressurized pipe or vessel containing a defect could be evaluated. It also shows how the extension of the defect at various lifetimes could be predicted. These relationships are prerequisite for the probabilistic formulation and analysis for the periodic inspection of piping systems

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

International Nuclear Information System (INIS)

Mohd Jamil Bin Hashim

2011-01-01

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

77 FR 27428 - Certain Large Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe (Over 41/2

Science.gov (United States)

2012-05-10

... (``ASME'') code stress levels. Alloy pipes made to ASTM A-335 standard must be used if temperatures and stress levels exceed those allowed for ASTM A-106. Seamless pressure pipes sold in the United States are..., or pressure pipe applications. B. Finished and unfinished oil country tubular goods (``OCTG''), if...

76 FR 66688 - Certain Large Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe (Over 41/2

Science.gov (United States)

2011-10-27

... (``ASME'') code stress levels. Alloy pipes made to ASTM A-335 standard must be used if temperatures and stress levels exceed those allowed for ASTM A-106. Seamless pressure pipes sold in the United States are..., or pressure pipe applications. B. Finished and unfinished oil country tubular goods (``OCTG''), if...

Safety technology qualification of the prestressed cast iron pressure vessel (PCIV) and of the primary cell of the HTR-modul for the passive removal of decay heat, phase 1 (INHR)

International Nuclear Information System (INIS)

Warnke, E.P.

1990-02-01

During this development program the thermodynamic behaviour of a system was investigated, consisting of a hot working Prestressed Cast Iron Pressure Vessel and an inactive heat sink in the surrounding cavern cell. It could be shown, that the inactive heat removal system designed as a natural circuit can remove the maximum amount of heat of 890 kW during emergency conditions via a natural-draught air cooling tower even under very conservative assumptions and for a 50% loss of cooling pipes. Further it could be shown, that the hot working Prestressed Cast Iron Pressure Vessel has a very safe load carrying behaviour during all normal and upset conditions. (orig.) With 10 tabs., 38 figs., 43 refs [de

Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

1998-07-01

SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

Method of detecting construction faults in concrete pressure vessels

International Nuclear Information System (INIS)

Robertson, S.A.; Duhoux, M.; Dawance, G.; Carrie, C.; Morel, D.

1976-01-01

A major problem in the design and construction of concrete pressure vessels for nuclear power stations is the risk of excessive air leaks through the concrete itself, due to faulty construction. The 'sonic coring' method of non-destructive concrete testing has been used successfully in pile and diaphragm wall construction control for several years, and the potential use of this method to control the presence of faults in concrete pressure vessels is here described. (author)

Cylindrical reinforced-concrete pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

Vaessen, F.

1975-01-01

The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

Rotor characteristics experiments with cooling pipe/fender bolt-groups, HTR-500

International Nuclear Information System (INIS)

Froening.

1990-01-01

For the design of composite liners for prestressed concrete pressure vessels of high temperature reactors, the force-displacement relationship of rotor elements is required. The rotor elements link the liner with the vessel concrete and transfer loads between liner and concrete. The liner is mainly stressed only at the sheet level due to concrete upsetting because of external load, including creeping due to prestress, internal pressure and temperature, shrinkage and temperature difference between liner and concrete. In the region of planned peripheral disturbances (penetrations, sheet/transfers, etc.), and at places with unintentional imperfections (dents, points of impact with slightly different sheet thicknesses), the liner sheet may be displaced against the concrete. Such displacements lead to a shear load on fender bolts and cooling pipes. The strains develop slowly. They can also be cyclical owing to internal pressure and temperature changes, the amplitudes being low, however. In addition, tensile stress may be applied to the fender bolt rotors by the driving effect of the shear-stressed cooling pipes, by the liner bending strain at the edges of dented fields, or by external loads. (orig.) [de

Pipe connection for high pressure and temperature loads

International Nuclear Information System (INIS)

Haferkamp, D.; Hodzic, A.; Paetz, E.; Stach, H.

1976-01-01

The patent proposes an inprovement of the clamping device for a pipe joint connecting pipelines which are subject to high pressure and temperature loads, e.g. in a nuclear power plant. This clamping device may be tightened and loosened by remote control. The proposed clamping ring consists of several segments connected with each other by hinge-type guide pins and fishplates. (UWI) [de

Design of pressure vessels. Part 1

International Nuclear Information System (INIS)

Grandemange, J.M.

2008-01-01

The equipments and loops of PWR reactors are basically pressure vessels. Their specificities concern the integrity warranties that must be implemented considering their importance for the reactors safety. Thus, stress is put on the exhaustiveness of the prevention of in-service degradation and on the safety scenarios considered. The second specificity concerns the possibility of activation of wear and corrosion products during their flow inside the reactor core. This second aspect leads to some constraints on the choice of the materials used and on the surface coating of the inside wall of big components of the primary circuit. The aim of this document is to develop the general approach adopted for the design of the pressure vessels of PWR fluid loops, and to stress more particularly on the nuclear particularities of these equipments. Some extensions of these rules to high temperature resistant materials (FBR-type reactors) are also evoked. Content: General considerations: design basis of pressure vessels, risk analysis and design conditions, ruining paths and safety coefficients; 2 - damage prevention for excessive deformation: definitions, criteria; 3 - prevention of the plastic instability damage: definition, criteria; 4 - buckling prevention: definition and mechanisms, rules and criteria; 5 - prevention of progressive deformation damage: definitions, plastic adaptation, plastic accommodation, progressive deformation; 6 - prevention of fatigue damage: definitions, general prevention approach, design fatigue curves, analytic approach, particular aspects, analysis of zones with geometrical singularity; 7 - prevention of sudden rupture damage: fragile rupture and ductile tear, general approach, analytic criteria, irradiation and aging effects; 8 - other potential damages; 9 - conclusion. (J.S.)

Determination of Burst Pressure of API Steel Pipes using Stress Modified Critical Strain Model

International Nuclear Information System (INIS)

Alang, N A; Razak, N A; Sulaiman, A S

2012-01-01

This paper presents a technique which can be used to determine the burst pressure of defective steel pipes using non-linear finite element (FE) analysis. The technique uses stress modified critical strain (SMCS) failure criterion to study the effect of gouge defects on maximum working pressure of API X65 steel pipes. The procedures in determining the model parameters using 3-D, homogeneous isotropic elastic-plastic material model with large deformation finite element analyses from notched tensile bars were systematically discussed. The relationship between burst pressure and gouge depth was proposed. The burst pressure estimated then was compared to experimental data from the literature for validation showing overall good agreements.

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Pajnic, M.

2006-01-01

The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

In-place thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1985-04-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

Investigation of the failure of a reactor pressure vessel by plastic instability

International Nuclear Information System (INIS)

Laemmer, H.; Ritter, B.

1994-01-01

A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

Ultrasonic stress evaluation through thickness of a stainless steel pressure vessel

International Nuclear Information System (INIS)

Javadi, Yashar; Pirzaman, Hamed Salimi; Raeisi, Mohammadreza Hadizadeh; Najafabadi, Mehdi Ahmadi

2014-01-01

This paper investigates ultrasonic method in stress measurement through thickness of a pressure vessel. Longitudinal critically refracted (L CR ) waves are employed to measure the welding residual stresses in a vessel constructed from austenitic stainless steel 304L. The acoustoelastic constant is measured through a hydro test to keep the pressure vessel intact. Hoop and axial residual stresses are evaluated by using different frequency range of ultrasonic transducers. The welding processes of vessel shell and caps are simulated by a 3D finite element (FE) model which is validated by hole-drilling method. The residual stresses calculated by FE simulation are then compared with those obtained from the ultrasonic measurement while a good agreement is observed. It is demonstrated that the residual stresses through thickness of the stainless steel pressure vessel can be evaluated by combining FE and L CR method (known as FEL CR method). - Highlights: • The main goal is ultrasonic evaluation of through thickness stresses. • Welding processes of a stainless steel pressure vessel are modelled by FE. • The hole-drilling method is used to validate the FE results. • Residual stresses are measured by four different series of ultrasonic transducers. • The comparison between ultrasonic and FE results show an acceptable agreement

A prestressed concrete pressure vessel for helium high temperature reactor system

International Nuclear Information System (INIS)

Horner, R.M.W.; Hodzic, A.

1976-01-01

A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

Large inelastic deformation analysis of steel pressure vessels at high temperature

International Nuclear Information System (INIS)

Ikonen, K.

2001-01-01

This publication describes the calculation methodology developed for a large inelastic deformation analysis of pressure vessels at high temperature. Continuum mechanical formulation related to a large deformation analysis is presented. Application of the constitutive equations is simplified when the evolution of stress and deformation state of an infinitesimal material element is considered in the directions of principal strains determined by the deformation during a finite time increment. A quantitative modelling of time dependent inelastic deformation is applied for reactor pressure vessel steels. Experimental data of uniaxial tensile, relaxation and creep tests performed at different laboratories for reactor pressure vessel steels are investigated and processed. An inelastic deformation rate model of strain hardening type is adopted. The model simulates well the axial tensile, relaxation and creep tests from room temperature to high temperature with only a few fitting parameters. The measurement data refined for the inelastic deformation rate model show useful information about inelastic deformation phenomena of reactor pressure vessel steels over a wide temperature range. The methodology and calculation process are validated by comparing the calculated results with measurements from experiments on small scale pressure vessels. A reasonably good agreement, when taking several uncertainties into account, is obtained between the measured and calculated results concerning deformation rate and failure location. (orig.)

Reliability analysis of pipelines and pressure vessels at nuclear power plants

International Nuclear Information System (INIS)

Klemin, A.I.; Shiverskij, E.A.

1979-01-01

Reliability analysis of pipelines and pressure vessels at NPP is given. The main causes and failure mechanisms of these elements, the ways of reliability improvement and preventing of great damages are considered. The reliability estimation methods both according to the statistical operation data and under the conditions of absence of failure statistics are given. The main characteristics and actual reliability factors of pipelines and pressure vessels of three home NPP: the first in the world NPP, VK-50 and Beloyarsk NPP, are presented. From the start-up there were practically no failures of the pipelines and pressure vessels at the VK-50 pilot installation. The analysis of the operation experience of the first and second blocks of the Beloyarsk NPP, as well as the first in the world NPP, shows that the most part of failures of the pipelines and pressure vessels of these energy blocks with the channel reactors is connected with the coolant leakage at minority pipelines of a small diameter. The most part of failures at individual pipelines of the first and second blocks of the Beloyarsk NPP are connected with the leakages of stuffing boxes of switching off devices. It is noted that serious failures of large pipelines and pressure vessels at all home NPP under operation have not been observed

Heritability of retinal vessel diameters and blood pressure

DEFF Research Database (Denmark)

Taarnhøj, Nina C B B; Larsen, Michael; Sander, Birgit

2006-01-01

PURPOSE: To assess the relative influence of genetic and environmental effects on retinal vessel diameters and blood pressure in healthy adults, as well as the possible genetic connection between these two characteristics. METHODS: In 55 monozygotic and 50 dizygotic same-sex healthy twin pairs......%-80%) for CRAE, 83% (95% CI: 73%-89%) for CRVE, and 61% (95% CI: 44%-73%) for mean arterial blood pressure (MABP). Retinal artery diameter decreased with increasing age and increasing arterial blood pressure. Mean vessel diameters in the population were 165.8 +/- 14.9 microm for CRAE, 246.2 +/- 17.7 microm...... for CRVE, and 0.67 +/- 0.05 microm for AVR. No significant influence on artery or vein diameters was found for gender, smoking, body mass index (BMI), total cholesterol, fasting blood glucose, or 2-hour oral glucose tolerance test values. CONCLUSIONS: In healthy young adults with normal blood pressure...

Testing of Full Scale Flight Qualified Kevlar Composite Overwrapped Pressure Vessels

Science.gov (United States)

Greene, Nathanael; Saulsberry, Regor; Yoder, Tommy; Forsyth, Brad; Thesken, John; Phoenix, Leigh

2007-01-01

Many decades ago NASA identified a need for low-mass pressure vessels for carrying various fluids aboard rockets, spacecraft, and satellites. A pressure vessel design known as the composite overwrapped pressure vessel (COPV) was identified to provide a weight savings over traditional single-material pressure vessels typically made of metal and this technology has been in use for space flight applications since the 1970's. A typical vessel design consisted of a thin liner material, typically a metal, overwrapped with a continuous fiber yarn impregnated with epoxy. Most designs were such that the overwrapped fiber would carry a majority of load at normal operating pressures. The weight advantage for a COPV versus a traditional singlematerial pressure vessel contributed to widespread use of COPVs by NASA, the military, and industry. This technology is currently used for personal breathing supply storage, fuel storage for auto and mass transport vehicles and for various space flight and aircraft applications. The NASA Engineering and Safety Center (NESC) was recently asked to review the operation of Kevlar 2 and carbon COPVs to ensure they are safely operated on NASA space flight vehicles. A request was made to evaluate the life remaining on the Kevlar COPVs used on the Space Shuttle for helium and nitrogen storage. This paper provides a review of Kevlar COPV testing relevant to the NESC assessment. Also discussed are some key findings, observations, and recommendations that may be applicable to the COPV user community. Questions raised during the investigations have revealed the need for testing to better understand the stress rupture life and age life of COPVs. The focus of this paper is to describe burst testing of Kevlar COPVs that has been completed as a part of an the effort to evaluate the effects of ageing and shelf life on full scale COPVs. The test articles evaluated in this discussion had a diameter of 22 inches for S/N 014 and 40 inches for S/N 011. The

Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

International Nuclear Information System (INIS)

Chiaraviglio, N.; Bazzana, S.

2013-01-01

Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

International Nuclear Information System (INIS)

Adamec, P.

2000-12-01

Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

Recent experiences and problems in conducting pressure vessel surveillance examinations

International Nuclear Information System (INIS)

Perrin, J.S.

1979-01-01

Each of the commercial power reactors in the U.S.A. has a pressure vessel surveillance program. The purpose of the programs is to monitor the effects of radiation on the mechanical properties on the steel pressure vessels. A program for a given reactor includes a series of irradiation capsules containing neutron dosimeters and mechanical property specimens. The capsules are periodically removed during the life of the reactor and evaluated. The surveillance capsule examinations conducted to date have been valuable in assessing the effects of radiation on pressure vessels. However, a number of problems have been observed in the course of capsule examinations which potentially could reduce the maximum value of the data obtained. These problems are related to specimen design and preparation, capsule design and preparation, capsule installation and removal, capsule disassembly, specimen testing and evaluation, program documentation, and quality assurance. Examples of problems encountered in the preceding areas are presented in the present paper, and recommendations are made for minimization or prevention of these problems in future programs. Included in the recommendations is that appropriate ASTM standards, ASME Boiler and Pressure Vessel Code sections, and NRC regulations provide the appropriate framework for prevention of problems

A quantitative methodology for reactor vessel pressurized thermal shock decision making

International Nuclear Information System (INIS)

Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

1983-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

Welding of the A1 reactor pressure vessel

International Nuclear Information System (INIS)

Becka, J.

1975-01-01

As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

East/west steels for reactor pressure vessels

International Nuclear Information System (INIS)

Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

1997-01-01

The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

46 CFR 32.50-15 - Cargo piping on tank vessels constructed on or after July 1, 1951-TB/ALL.

Science.gov (United States)

2010-10-01

... TANK VESSELS SPECIAL EQUIPMENT, MACHINERY, AND HULL REQUIREMENTS Pumps, Piping, and Hose for Cargo... heavy, all joints are welded, and bends are installed to provide for expansion and contraction. (2) Tank... shall not pass through spaces containing machinery where sources of vapor ignition are normally present...

Effect of Operating Pressure on Hydrogen Risk in Filtered Containment Venting System

Energy Technology Data Exchange (ETDEWEB)

Na, Young Su; Cho, Song-Won; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The FCVS (Filtered Containment Venting System) has the main objectives of both the depressurization in the containment building and the decontamination of fission products generated under a severe accident. One of the commercial wet-type FCVSs consists of a cylindrical pressure vessel including a scrubbing solution and filters. A FCVS vessel can be installed on the outside of the containment building, and is connected with the containment through a pipe. When the pressure in the containment building approaches the setting value, a valve on a pipe between the containment and the FCVS opens to operate the FCVS. The amount of steam and gas mixtures generated under a severe accident can be released into the FCVS, where the nozzles of a pipe are submerged into a scrubbing solution in a FCVS vessel. Non-condensable gases and fine aerosols can enter a scrubbing solution, and they then pass the filters. The decontaminated gases are finally discharged from the FCVS into the outside environment. Previous studies have introduced critical issues with the operation of the FCVS. Reference [2] assessed the effect of the operating pressure of the FCVS on the hydrogen risk in a FCVS vessel. The volumetric concentrations of hydrogen and steam in a postulated FCVS with a 3 m diameter and 6.5 m height were calculated using the MELCOR computer code (v. 1.8.6). After the operation of the FCVS, the pressure and temperature in the FCVS vessel jumped from the initial conditions of the atmosphere pressure and room temperature. For the FCVS operating pressure of 5 bar, the hydrogen concentration increased from 6% in the containment to 14% in a FCVS vessel, whereas the steam concentration decreased from 58% in the containment to 3% in a FCVS vessel. The increased hydrogen concentration with air in a FCVS vessel can exists within the region of the burn limit in the Shapiro diagram. This possibility of the hydrogen combustion can threaten the integrity of the FCVS. To mitigate the hydrogen risk

Pressure vessel inspection criteria based on fitness-for-purpose assessment

International Nuclear Information System (INIS)

Grover, J.L.; Cipolla, R.C.

1985-01-01

The paper on pressure vessel inspection investigates the methodology required to establish an inspection strategy consistent with fracture mechanics analysis, i.e. to define allowable flaw sizes based on location within the vessel. The methodology is demonstrated using a sample problem for a typical pressurised water reactor pressure vessel, and shows the impact of certain assumptions on the inspection strategy. The results indicate that the flaw size varies with the shape of the assumed residual stress field and the through-thickness location. Also in general, the fracture mechanics evaluation allows flaws much larger than are allowed by the inspection acceptance criteria. (UK)

Development of automatic ultrasonic testing equipment for reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-08-01

The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author).

Development of automatic ultrasonic testing equipment for reactor pressure vessel

International Nuclear Information System (INIS)

Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik

1994-08-01

The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author)

77 FR 21734 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Romania...

Science.gov (United States)

2012-04-11

... DEPARTMENT OF COMMERCE International Trade Administration [A-485-805] Certain Small Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Romania: Extension of Time Limit for... diameter carbon and alloy seamless standard, line and pressure pipe from Romania for the period August 1...

Multilayer Pressure Vessel Materials Testing and Analysis. Phase 1

Science.gov (United States)

Cardinal, Joseph W.; Popelar, Carl F.; Page, Richard A.

2014-01-01

To provide NASA a comprehensive suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for aging multilayer pressure vessels, Southwest Research Institute (R) (SwRI) was contracted in two phases to obtain relevant material property data from a representative vessel. This report describes Phase 1 of this effort which includes a preliminary material property assessment as well as a fractographic, fracture mechanics and fatigue crack growth analyses of an induced flaw in the outer shell of a representative multilayer vessel that was subjected to cyclic pressure test. SwRI performed this Phase 1 effort under contract to the Digital Wave Corporation in support of their contract to Jacobs ATOM for the NASA Ames Research Center.

Prediction of Composite Pressure Vessel Failure Location using Fiber Bragg Grating Sensors

Science.gov (United States)

Kreger, Steven T.; Taylor, F. Tad; Ortyl, Nicholas E.; Grant, Joseph

2006-01-01

Ten composite pressure vessels were instrumented with fiber Bragg grating sensors in order to assess the strain levels of the vessel under various loading conditions. This paper and presentation will discuss the testing methodology, the test results, compare the testing results to the analytical model, and present a possible methodology for predicting the failure location and strain level of composite pressure vessels.

Experimental and theoretical studies of transient boiling and two-phase flow during the depressurisation of a simple glass vessel

International Nuclear Information System (INIS)

Ardron, K.H.; Furness, R.A.; Hall, P.C.

1976-11-01

Blowdown experiments using a glass pressure vessel containing saturated water at 4 bars have been performed to assist interpretation of the results of large scale experiments and aid understanding of the physical processes involved. Results have shown the strong dependence of depressurisation time, phase distribution and mass flow rate on the length to diameter ratio of the exit pipe. Preliminary observations of the flow regime in the discharge pipe are consistent with predictions of the flow regime map of Mandhane, Gregory and Aziz 1974. Different flow regimes have been observed at different axial positions along the pipe. Bubble growth rates during the non-equilibrium phase of blowdown are shown to be in reasonable agreement with a simple convective heat flux analysis previously used in blowdown calculations. The transient pressure and liquid distribution in the vessel have been compared with calculations using the blowdown code RELAP-UK. (U.K.)

Rupture tests with reactor pressure vessel head models

International Nuclear Information System (INIS)

Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

2003-01-01

In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

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

Directory of Open Access Journals (Sweden)

Guoquan Qi

2015-11-01

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

Numerical evaluation of stress intensity factor for vessel and pipe subjected to thermal shock

International Nuclear Information System (INIS)

Kim, Y.W.; Lee, H.Y.; Yoo, B.

1994-01-01

The thermal weight function method and the finite element method were employed in the numerical computation of the stress intensity factor for a cracked vessel and the cracked pipe subjected to thermal shock. A wall subjected to thermal shock was analyzed, and it has been shown that the effect of thermal shock on the stress intensity factor is dominant for the crack with small crack length to thickness ratio. Convection at the crack face had an influence on the stress intensity factor in the early stage of thermal shock. (Author)

Application of Non-pressure Reinforced Concrete Pipes in Modern Construction and Reconstruction of Highways

Science.gov (United States)

Rakitin, B. A.; Pogorelov, S. N.; Kolmogorova, A. O.

2017-11-01

Modern highway construction technologies provide for the quality water discharge systems to increase facilities’ service life. Pipeline operating conditions require the use of durable and reliable materials and structures. The experience in using reinforced concrete pipes for these purposes shows their utilization efficiency. The present paper considers the experience in the use of non-pressure reinforced concrete pipes manufactured by the German company SCHLOSSER-PFEIFFER under the Ural region geological and climatic conditions. The authors analyzed the actual operation of underground pipelines and effective loads upon them. A detailed study of the mechanical properties of reinforced concrete pipes is necessary to improve their production technology and to enhance their serviceability. The use of software-based methods helped to develop a mathematical model and to estimate the strength and crack resistance of reinforced concrete pipes at different laying depths. The authors carried out their complex research of the strain-stress behaviour of reinforced concrete pipes and identified the most hazardous sections in the structure. The calculations performed were confirmed by the results of laboratory tests completed in the construction materials, goods, and structures test center. Based on the completed research, the authors formulated their recommendations to improve the design and technology of non-pressure reinforced concrete pipes.

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

Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

International Nuclear Information System (INIS)

Neunert, B.; Jueptner, G.; Kumpf, H.

1975-01-01

The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

In-service inspection program for the NCS-80 reactor pressure vessel

International Nuclear Information System (INIS)

Scharge, J.; Wehowsky, P.; Zeibig, H.

1978-01-01

The in-service inspection program of reactor pressure vessels is mainly based on the ultra-sonic method, visual checking of inner and outer surfaces as well as pressure and leak tests. The test procedure require a design of the pressure vessel suitable for the test methods and the possibility to remove the pressure vessel internals. For the outside inspection a gap of sufficient width is mandatory. The present status of the ultra-sonic method and of the inner and outer manipulators affords to conduct the in-service inspection program in form of automatic checkings. The in-service inspection program for NCS-80, the Nuclear Container-Ship design of 80,000 shp, is integrated in the refueling periods due to the request for a high availability of the ship and reactor plant

Design Improvement of Double Pressure Vessel in the In-pile Test Section

Energy Technology Data Exchange (ETDEWEB)

Hong, Jintae; Heo, Sung-Ho; Joung, Chang-Young; Kim, Ka-Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

Initiation and arrest - two approaches to pressure vessel safety

International Nuclear Information System (INIS)

Brumovsky, M.; Filip, R.; Stepanek, S.

1976-01-01

The safety analysis is described of the reactor pressure vessel related to brittle fracture based on the fracture mechanics theory using two different approximations, i.e., the Crack Arrest Temperature (CAT) or Nil Ductility Temperature (NDT), and fracture toughness. The variation of CAT with stress was determined for different steel specimens of 120 to 200 mm in thickness. A diagram is shown of CAT variation with stress allowing the determination of crack arrest temperature for all types of commonly used steels independently of the NDT initial value. The diagram also shows that the difference between fracture transition elastic (FTE) and NDT depends on the type of material and determines the value of the ΔTsub(sigma) factor typical of the safety coefficient. The so-called fracture toughness reference value Ksub(IR) is recommended for the computation of pressure vessel criticality. Also shown is a defect analysis diagram which may be used for the calculation of pressure vessel safety prior to and during operation and which may also be used in making the decision on what crack sizes are critical, what cracks may be arrested and what cracks are likely to expand. The diagram is also important for the fact that it is material-independent and may be employed for the estimates of pre-operational and operational inspections and for pressure vessel life prediction. It is generally applicable to materials of greater thickness in the region where the validity of linear elastic fracture mechanics is guaranteed. (J.P.)

Manufacturing and use of spiral welded pipes for high pressure service : state of the art

Energy Technology Data Exchange (ETDEWEB)

Knoop, F.M.; Sommer, B. [Salzgitter GroBrohre GmbH, Salzgitter (Germany)

2004-07-01

This paper provided details of an improved helical seam 2-step (HTS) manufacturing process used to produce spiral welded large diameter pipes for high pressure transmission pipelines. During the process, pipe forming is combined with continuous tack welding and internal and external submerged arc welding at separate welding stations. The pipe forming unit consists of a 3 roll bending system with an outside roller cage used to guarantee the roundness of the pipe. The converging strip edges of the pipe are joined using a continuous shielded arc tack weld. Tack welding is done automatically with a laser-guided weld head. Run-out angles are adjusted by an automatic gap control system. The formed and tack-welded pipes are then fed to computer-controlled welding stations for final welding, where each pipe is rotated with a precise screw-like motion. The same welding materials used for the helical seam are used for the skelp end welding. The process offers more precise root gap control, as well as improved pipe geometry. Use of the process has also increased production rates and improved weld stability. The dimensions of the spiral-weld pipes are adjustable so that any diameter can be produced from a base material of the same width. The pipes can also be coated externally with fusion-bonded epoxy or 3-layer polyethylene/polypropylene. It was concluded that the process is being further refined to support the use of HTS pipes in high-pressure pipelines. New nondestructive testing techniques used to assess the performance of the line pipes were presented, as well as the results from hot and cold bending tests, field weldability trials, and tests related to the safety of spiral pipes. 16 refs., 2 tabs., 12 figs.

Phenomenological vessel burst investigations

International Nuclear Information System (INIS)

Hippelein, K.W.; Julisch, P.; Muz, J.; Schiedermaier, J.

1985-07-01

Fourteen burst experiments have been carried out using vessels with circumferential and longitudinal flaws, for investigation of the fracture behaviour, i.e. the time-related fracture opening. The vessels had dimensions (outer diameter x wall thickness = 800 x 47 mm) which correspond to the dimensions of the main coolant piping of a 1300 MW e PWR. The test specimens had been made of the base-safe material 20 MnMoNi 55 and of a special, 22 NiMoCr 37 base alloy. The experimental conditions with regard to pressure and temperature have been chosen so as to correspond to normal operating conditions of a PWR (p∝17.5 MPa, T∝300 0 C), i.e. the flaws have been so dimensioned that failure was to be expected at a pressure of p∝17.5 MPa. As a rule, water has been used as the pressure medium, or in some cases air, in order to influence the time-dependent pressure decrease. Fluid and structural dynamics calculations have also been made. In order to determine the impact of a fast propagating crack on the leak-to-fracture curve, which normally is defined by quasistationary experiments, suitable tests have been made with large-volume, cylindrical vessels (outer diameter x wall thickness x length = 3000 x 21 x 14000 mm) made of the material WSt E 43. The leak-before-fracture criterion has been confirmed. (orig./HP) [de

An introduction to the analysis of multi-cavity prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Silva, M.C.A.T. da.

1986-01-01

The present work is a study of multi-cavity prestressed concrete pressure vessels (PCRV) for nuclear reactors. A review is made of the designs, analises and models of multi-cavity concrete pressure vessels. A preliminary evaluation of the NONSAP program for applications in complex three-dimensional structures such as a multi-cavity pressure vessel is also made. A model of a PCRV of a 1000 MW(e) high-temperature gas cooled reactor was selected for a three-dimensional analysis with the NONSAP program. The results obtained are compared with experimental data. (Author) [pt

Analysis of cracked pressure vessel nozzles by finite elements

International Nuclear Information System (INIS)

Reynen, J.

1975-01-01

In order to assess the safety of pressure vessel nozzles, the analysis should take into account cracks. The paper describes various algorithms, their computer implementations and relative merits to define in an effective way strain energy release rates along the tip front of arbitrary 3 D cracks under arbitary load including thermal strains. These techniques are basically equivalent to substructuring techniques and consequently they can be implemented to only FEM program able to deal with the data handling problems of the substructuring technique. Examples are given carried out with a substructure version of the BERSAFE system. These examples include a corner crack in a pressure vessel nozzle loaded by internal pressure and by thermal stresses. (Auth.)

Helium leak testing of large pressure vessels or subassemblies

International Nuclear Information System (INIS)

Hopkins, J.S.; Valania, J.J.

1977-01-01

Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994

Energy Technology Data Exchange (ETDEWEB)

Lowendahl, B. [OKG Aktiebolag, Figeholm (Sweden)

1995-03-01

In September 1992 Oskarshamn 1 was shut down in order to carry out measures to correct discovered deficiencies in the emergency cooling systems. Due to the results of a comprehensive non destructive test programme it was decided to perform a major replacement of pipes in the primary systems including a full system decontamination using the Siemens CORD process. The paper briefly presents the satisfying result of the decontamination performed in May-June 1993. When in late June 1993 cracks also were detected in the feed-water pipes situated inside the reactor pressure vessel (RPV) the plans were reconsidered and a large project was formed with the aim, in a first phase, to verify the integrity of the RPV. In order to make it possible to perform work manually inside the RPV special radiation protection measures had to be carried out. In January 1994 the lower region of the RPV was decontaminated, again using the CORD-process, followed by the installation of a special shielding construction in the RPV. The surprisingly good results of these efforts are also briefly described in the paper.

A mathematical model for pressure-based organs behaving as biological pressure vessels.

Science.gov (United States)

Casha, Aaron R; Camilleri, Liberato; Gauci, Marilyn; Gatt, Ruben; Sladden, David; Chetcuti, Stanley; Grima, Joseph N

2018-04-26

We introduce a mathematical model that describes the allometry of physical characteristics of hollow organs behaving as pressure vessels based on the physics of ideal pressure vessels. The model was validated by studying parameters such as body and organ mass, systolic and diastolic pressures, internal and external dimensions, pressurization energy and organ energy output measurements of pressure-based organs in a wide range of mammals and birds. Seven rules were derived that govern amongst others, lack of size efficiency on scaling to larger organ sizes, matching organ size in the same species, equal relative efficiency in pressurization energy across species and direct size matching between organ mass and mass of contents. The lung, heart and bladder follow these predicted theoretical relationships with a similar relative efficiency across various mammalian and avian species; an exception is cardiac output in mammals with a mass exceeding 10kg. This may limit massive body size in mammals, breaking Cope's rule that populations evolve to increase in body size over time. Such a limit was not found in large flightless birds exceeding 100kg, leading to speculation about unlimited dinosaur size should dinosaurs carry avian-like cardiac characteristics. Copyright © 2018. Published by Elsevier Ltd.

Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

2009-01-01

The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

Assessment of the integrity of WWER type reactor pressure vessels

International Nuclear Information System (INIS)

Brumovsky, M.

1995-01-01

Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

Stochastic simulation of PWR vessel integrity for pressurized thermal shock conditions

International Nuclear Information System (INIS)

Jackson, P.S.; Moelling, D.S.

1984-01-01

A stochastic simulation methodology is presented for performing probabilistic analyses of Pressurized Water Reactor vessel integrity. Application of the methodology to vessel-specific integrity analyses is described in the context of Pressurized Thermal Shock (PTS) conditions. A Bayesian method is described for developing vessel-specific models of the density of undetected volumetric flaws from ultrasonic inservice inspection results. Uncertainty limits on the probabilistic results due to sampling errors are determined from the results of the stochastic simulation. An example is provided to illustrate the methodology

Fabrication techniques of metal liner used for pressure vessels made by composite material

International Nuclear Information System (INIS)

Takahashi, W.K.; Al-Qureshi, H.A.

1982-01-01

Different viable techniques for the manufacturing of metal liner used for pressure vessels are presented. The aim of these metal liner is to avoid the fluid leakage from the pressurized vessel and to serve as a mandreal to be wound by composite material. The studied techniques are described and the practical results are illustrated. Finally a comparative study of the manufacturing techniques is made in order to define the process that furnishes the metal liner with the best characteristics. The advantages offered by these type of pressure vessels when compared with the conventional metallic vessels, are also presented. (Author) [pt

Test of 6-inch-thick pressure vessels. Series 2. Intermediate test vessels V-3, V-4, and V-6

International Nuclear Information System (INIS)

Bryan, R.H.; Merkle, J.G.; Raftenberg, M.N.; Robinson, G.C.; Smith, J.E.

1975-11-01

The second series of intermediate vessel tests were crack initiation fracture tests of 6-in.-thick 39-in.-OD steel vessels with sharp surface flaws approximately 2 1 / 2 in. deep by 8 in. long in the longitudinal weld seams of the test cylinders. Fracture was initiated by means of hydraulic pressurization. One vessel was tested at each of three temperatures: 75, 130, and 190 0 F. Pretest analyses were made to predict the failure pressures and strains. Fracture toughness data obtained by equivalent-energy analysis of precracked Charpy-V tests and compact-tension specimen tests were used in the fracture analyses. The vessels behaved generally as had been expected. Posttest fracture analyses were also performed for each vessel. Detailed discussions of the fracture analysis methods developed in support of the vessel tests described are included. 34 references