46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

Science.gov (United States)

2010-10-01

...) NAUTICAL SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-5 Inspection of boilers, pressure vessels, piping and appurtenances. The inspection of boilers, pressure vessels, piping and appurtenances... 46 Shipping 7 2010-10-01 2010-10-01 false Inspection of boilers, pressure vessels, piping and...

Validated automated ultrasonic inspections of the Sizewell 'B' reactor pressure vessel

International Nuclear Information System (INIS)

Dikstra, B.J.; Farley, J.M.

1992-01-01

Automated ultrasonic inspection was applied extensively during manufacture of the RPV for Sizewell 'B'. This was an important element of the safety case presented at the Sizewell 'B' public enquiry. This requirement reflected concern in the United Kingdom as to the effectiveness and reliability of ultrasonic inspections. By applying automated inspections in addition to the manual ultrasonic inspection carried out by the vessel manufacturer, the overall reliability of the inspection of the vessel would be considerably enhanced. The automated inspections carried out in the manufacturer's workshops were termed 'automated shop inspections' (ASIs). The ASIs were carried out in two contracts: the first to inspect the component forgings of the RPV, the second to inspect the pressure retaining welds. (author)

Inter-vessels in-service inspection of Super-Phenix

International Nuclear Information System (INIS)

Asty, M.; Saglio, R.; Viard, J.; Lerat, B.

1984-01-01

The vessels design of fast breeder reactor Super-Phenix enables inspection during operating time. A self-moving machine -MIR- has been built up especially for that purpose. It is able to carry out visual and ultrasonorous inspection. MIR structure is that of a tetrahedron, all tops of which are fitted with two wheels, as for traction and direction. The wheels are leaning on booth the two vessels. Thanks to a computer-assisted control system, MIR is able to move along in every part of the inter-vessels space. Studies have been carried on at the French Commissariat a l'Energie Atomique, by two Sections of the advanced technologies Service. After outlining MIR working conditions, its main characteristics are described [fr

Qualification of UT methods and systems used for in-service inspections of VVER 440 vessels

International Nuclear Information System (INIS)

Skala, Z.; Vit, J.

2003-01-01

SKODA JS has been performing automated in-service inspections VVER reactor pressure vessels for more than twenty years. All of these inspections were performed by ultrasonic pulse echo method, combined from 1996 with eddy current testing. The Time of Flight Diffraction Method (TOFD) is one of modern methods of ultrasonic testing. The accuracy of sizing the through wall extent of a flaw by TOFD is much better than the accuracy achievable by the pulse echo method. A series of laboratory tests were performed by SKODA JS and confirmed the suitability of TOFD method for VVER reactor parts testing. The Czech Atomic law demands the qualification of systems and methods used for the in-service inspections of nuclear reactors. The qualification is done in accordance with ENIQ methodology and consists of preparation of the Technical Justification and practical tests made under the surveillance of Qualification Body. SKODA JS intends to qualify systems and methods used for the automated ultrasonic testing of VVER 440 and VVER 1000 reactor components from the inner as well as from the outer surface. The accuracy of the flaw through wall extent sizing by TOFD was confirmed by the qualification of methods and systems used for the testing of VVER 440 vessel circumferential weld and so the TOFD method shall be used routinely by SKODA JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level. (author)

75 FR 56015 - Vessel Inspection Alternatives

Science.gov (United States)

2010-09-15

... DEPARTMENT OF HOMELAND SECURITY Coast Guard 46 CFR Part 8 Vessel Inspection Alternatives CFR... Certificate; (ii) International Tonnage Certificate; (iii) Cargo Ship Safety Construction Certificate; (iv) Cargo Ship Safety Equipment Certificate; and (v) International Oil Pollution Prevention Certificate; and...

RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

International Nuclear Information System (INIS)

Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.

1984-01-01

Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

MIR: an in-service inspection device for Superphenix 1 vessels

International Nuclear Information System (INIS)

Asty, M.; Ceccato, S.; Lerat, B.; Viard, J.

1986-06-01

The main and safety vessels of SUPERPHENIX 1 were designed to allow in-service inspections. The remote controlled inspection device MIR was developed for this purpose. It allows both visual and ultrasonic examinations to be performed. Basically, MIR consists of a tetrahedral structure provided with four steering and traction wheels, two for each vessel. A computer assisted control system enables it to be driven to any position on either the main or safety vessels. Operating conditions are briefly reviewed and the main features of MIR presented

In-service inspection robot for PFBR main vessel- concept

Energy Technology Data Exchange (ETDEWEB)

Rajendran, S; Ramakumar, M S [Bhabha Atomic Research Centre, Mumbai (India). Div. of Remote Handling and Robotics

1994-12-31

In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs.

In-service inspection robot for PFBR main vessel- concept

International Nuclear Information System (INIS)

Rajendran, S.; Ramakumar, M.S.

1994-01-01

In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs

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

In service inspection of superphenix 1 vessels: MIR

International Nuclear Information System (INIS)

Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

1985-02-01

Presentation of the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of Super Phenix 1 vessels (surface and internal defects). The inspections take place during fuel handling operations. The inspection device is a robot with a four-wheel drive vehicle which guidance along the welds is achieved by eddy-current devices; visual examination is performed by a television camera and ultrasonic probes are specially resistent to high temperatures

Development of Ultrasonic Visual Inspection Program for In-Vessel Structures of SFR

International Nuclear Information System (INIS)

Joo, Y. S.; Park, C. G.; Lee, J. H.

2009-02-01

As the liquid sodium of a sodium-cooled fast reactor (SFR) is opaque to light, a conventional visual inspection is unavailable for the evaluation of the in-vessel structures under a sodium level. ASME Section XI Division 3 provides rules and guidelines for an in-service inspection (ISI) and testing of the components of SFR. For the ISI of in-vessel structures, the ASME code specifies visual examinations. An ultrasonic wave should be applied for an under-sodium visual inspection of the in-vessel structures. The plate-type waveguide sensor has been developed and the feasibility of the waveguide sensor technique has been successfully demonstrated for an ultrasonic visual inspection of the in-vessel structures of SFR. In this study, the C-scan image mapping program (Under-Sodium MultiView) is developed to apply this waveguide sensor technology to an under-sodium visual inspection of in-vessel structures in SFR by using a LabVIEW graphical programming language. The Under-Sodium MultiVIEW program has the functions of a double rotating scanner motion control, a high power pulser receiver control, a image mapping and a signal processing. The performance of Under-Sodium MultiVIEW program was verified by a C-scanning test

Inspection accessibility study of the Millstones Unit 1 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Calhoun, G.; Kapoor, A.; Davis, J.B.

1990-07-01

The need for more extensive ultrasonic examination of the Reactor Pressure Vessel (RPV) shell welds at some of the BWR plants is becoming apparent. Recent NRC comments regarding limits on future inspection relief requests and proposed revisions to the ASME Boiler and Pressure Vessel Code, Section 11 have caused BWR plant owners and vendors of inspection services to look at inspection tooling that would allow for the ultrasonic examination of RPV shell welds previously considered inaccessible. One approach to inspection of the RPV shell welds that will allow for greater coverage at some BWR plants is to perform these inspections from inside the vessel. In late 1988, Westinghouse received a contract from Northeast Utilities to examine the flange ligament areas and the shell welds in the upper portion of the Millstone Unit 1 PRV during an outage scheduled to begin in April 1989. This examination was performed with an inspection tool placed inside of the vessel and the UDRPS data acquisition system. The inspection tool was based on tooling used routinely at the Swedish and Finnish BWRs. The object of RP C105-1 was to use the information gathered prior to and during the Millstone Unit 1 examination and to perform an inspection accessibility study and prepare a report covering the lessons learned during the examination that would benefit other BWR utilities considering similar inspections. The final report consists of two volumes. This document, Volume 1, describes a 3D model of the Millstone Unit 1 PRV that was based on information obtained from plant drawings. The model is a useful tool for visualizing areas of limited access from both inside and outside of the RPV. 5 refs., 38 figs., 4 tabs

Recent development for inservice inspection of reactor pressure vessels

International Nuclear Information System (INIS)

Fischer, K.; Engl, G.; Rathgeb, W.; Heumueller, R.

1991-01-01

The German Nuclear Code (KTA-rules) requires a full scope inservice inspection (ISI) of reactor pressure vessels within a period of four years. This has a remarkable influence on plant operation and economy. Therefore, the development of advanced inspection equipment and techniques is directed not only to the enhancement of defect detectability and flaw sizing capabilities but also to reducing inspection times. A new manipulator system for PWR vessels together with fast data processing reduces the time for ISI of modern RPVs to 7 days. A new multichannel UT-system based on ALOK principle offers increased ultrasonic information with comfortable and rapid evaluation and presentation of results together with enhanced sizing capabilities. For specific inspection problems characterized by geometrical complexity the application of phased array probes in connection with UT-tomography provides improved ultrasonic information together with a streamlined manipulator principle and simplification of set up and tear down at the component which results in considerable reduction of radiation exposure. (orig.)

Programmable - logic equipment for ultrasound periodic inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Haniger, L.

1980-01-01

Two alternatives are presented of programmable logic corresponding to the 2nd generation of the apparatus for performing periodic ultrasonic inspections of power reactor pressure vessels and a solution is outlined of inspecting the circumferential weld on the pressure vessel head. The apparatus will allow using any measuring head taken into consideration for operational inspection. Command words are taken from a punched type reader. Czechoslovak made RAM memories are used. The algorithm of instrument function is supposed to be controlled by a microprocessor as soon as necessary preconditions for this technology are created in Czechoslovakia

Comparison of API 510 pressure vessels inspection planning with API 581 risk-based inspection planning approaches

International Nuclear Information System (INIS)

Shishesaz, Mohammad Reza; Nazarnezhad Bajestani, Mohammad; Hashemi, Seyed Javad; Shekari, Elahe

2013-01-01

To ensure mechanical integrity, all pressure vessels shall be inspected at the intervals provided in inspection codes or based on a risk-based inspection (RBI) assessment. The RBI assessment may allow previously established inspection intervals to be extended. This paper describes the methodology, analysis and results of two RBI studies conducted on 293 pressure vessel components in two crude oil distillation units. Based on API RBI methodology in API 581 (2008), risk target concept was used for determining inspection dates. It was shown that when thinning is the major active damage, the RBI recommended intervals are as long as twice the API 510 intervals. This paper summarizes that, as a fundamental step in the risk calculation, RBI has a more defined methodology for evaluating equipment for multiple damage mechanisms and a more defined approach to specify the use of other inspection technologies beyond the traditional visual, ultrasonic, and radiography tests. -- Highlights: • RBI calculated inspection intervals are as long as twice of API 510 inspection code. • Two case studies verified the advantage of RBI in inspection planning. • RBI is a more reliable methodology when evaluating multiple damage mechanisms. • Damage factor calculations can be used for determining RSFa value in FFS assessments

Inspection of dissimilar metal welds in reactor pressure vessels in Spain

Energy Technology Data Exchange (ETDEWEB)

Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P. [Tecnatom, S.A., San Sebastian de los Reyes, Madrid (Spain)

2011-07-01

MRP-139 recommendations for inspection of dissimilar metal (DM) welds in PWR vessels were launched in the last years in the USA. Basically, it increases the frequency of the examinations in these type of welds, with major emphasis in the hot loops, adding one intermediate inspection at the ten years interval in outlet nozzles. The spanish nuclear power plants (NPP's) have begun the implementation of this type of inspections on the vessel nozzles DM welds. As this type of inspections could have an impact in the critical path duration of the outage, it is necessary the use of a mechanical equipment able to examine the nozzles DM welds in a short vessel occupation time (VOT) with high quality, qualified techniques and minimum requirements of the refuelling platform. Tecnatom undertook the design and development of a new more advanced equipment, named TENIS-DM, for implementing the reactor pressure vessel (RPV) nozzles examination. This equipment was designed in order to accomplish the stringent requirements and the updated examination techniques; it was used for the inspection of the DM welds of Asco 1 NPP inlet and outlet nozzles in March 2011. Examination techniques and procedures were qualified through the GRUVAL validation program, based on ENIC methodology. Mechanical scanner was equipped with a large number of examination probes, and TV cameras -for visual inspection and also for monitoring the ultrasonic inspections. A remote operated submarine was also used to give support to the operational personnel during the manipulation of the equipment and its movements from one nozzle to the others. During two months before the inspection, tests of the complete inspection system were made on a nozzle mock-up installed in a 4 meters deep well at Tecnatom's facilities; this scenario was also used during the training sessions of the inspection crew. The defined technical and practical objectives were achieved: use of qualified techniques and minimal impact on the

Inspection of dissimilar metal welds in reactor pressure vessels in Spain

International Nuclear Information System (INIS)

Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P.

2011-01-01

MRP-139 recommendations for inspection of dissimilar metal (DM) welds in PWR vessels were launched in the last years in the USA. Basically, it increases the frequency of the examinations in these type of welds, with major emphasis in the hot loops, adding one intermediate inspection at the ten years interval in outlet nozzles. The spanish nuclear power plants (NPP's) have begun the implementation of this type of inspections on the vessel nozzles DM welds. As this type of inspections could have an impact in the critical path duration of the outage, it is necessary the use of a mechanical equipment able to examine the nozzles DM welds in a short vessel occupation time (VOT) with high quality, qualified techniques and minimum requirements of the refuelling platform. Tecnatom undertook the design and development of a new more advanced equipment, named TENIS-DM, for implementing the reactor pressure vessel (RPV) nozzles examination. This equipment was designed in order to accomplish the stringent requirements and the updated examination techniques; it was used for the inspection of the DM welds of Asco 1 NPP inlet and outlet nozzles in March 2011. Examination techniques and procedures were qualified through the GRUVAL validation program, based on ENIC methodology. Mechanical scanner was equipped with a large number of examination probes, and TV cameras -for visual inspection and also for monitoring the ultrasonic inspections. A remote operated submarine was also used to give support to the operational personnel during the manipulation of the equipment and its movements from one nozzle to the others. During two months before the inspection, tests of the complete inspection system were made on a nozzle mock-up installed in a 4 meters deep well at Tecnatom's facilities; this scenario was also used during the training sessions of the inspection crew. The defined technical and practical objectives were achieved: use of qualified techniques and minimal impact on the critical

INETEC new system for inspection of PWR reactor pressure vessel head

International Nuclear Information System (INIS)

Nadinic, B.; Postruzin, Z.

2004-01-01

INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

Apparative developments for inservice inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Bohn, H.; Ruthrof, K.; Barbian, O.A.; Kappes, W.; Neumann, R.; Stanger, H.K.

1987-01-01

Emphasizing PWR pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implenentation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system. (orig./HP)

Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

International Nuclear Information System (INIS)

Burns, D.C.

1979-01-01

Positioning means for locating inspection apparatus used to volumetrically examine a nuclear reactor vessel is disclosed. The positioning means is provided with a support ring having an annular key positioned longitudinally about its periphery. Three support legs are attached to the support ring by brackets adapted to fit the annular key. The support ring also carries three guide stud bushings which are movably mounted by clamps adapted to engage the support ring key. Prior to lowering the inspection apparatus into the vessel, the guide stud bushings are each moved to a point of alignment with one of three guide studs extending upwardly from the vessel. After alignment has been verified, the guide stud bushings are clamped in position. The inspection apparatus is lowered towards its fully seated position within the vessel and is coarsely circumferentially positioned with by the engagement of the guide studs within the guide stud bushings. A fine degree of circumferential positioning is achieved by providing a specially configured shoe for one of the support legs. With the core barrel internals in, the special shoe is adapted to key onto a core barrel pin the exact location of which is known. With the core barrel internals removed, the special shoe is adapted to place a locating key into a notch in a vessel flange, the location of which is known. As the inspection apparatus is lowered into its fully seated position, exact circumferential positioning with respect to the vessel is achieved. The other support legs rest on an inner circumferential flange so that no portion of the inspection apparatus touches or threatens the vessel's top flange. 19 claims

Remote controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Mueller, G.

1990-01-01

The first mechanised in-service inspection of the reactor pressure vessel on unit one of Eskom's Koeberg nuclear power station has been carried out. Since 1968 a whole range of manipulators to carry out remote controlled ultrasonic inspections of nuclear power station equipment has been developed. The inspection of a reactor pressure vessel using a central mast manipulator is described. 3 figs., 1 ill

Outlines of guidelines for the inspection and evaluation of reactor vessel internals

International Nuclear Information System (INIS)

Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

2014-01-01

'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

International Nuclear Information System (INIS)

Wang, Hesheng; Xu, Lifei; Chen, Weidong

2016-01-01

In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng; Xu, Lifei [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China)

2016-05-15

In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

In-service inspection of nuclear reactor vessels and steam generators. Results and evolution of the technics

International Nuclear Information System (INIS)

Rapin, Michel; Saglio, Robert.

1978-01-01

Methods and original technics have been developed by the CEA for inspection of the primary coolant circuit of PWR. Multifrequency Eddy currents for inspection of steam generators tubes gudgeons and bolts; focussed ultrasonics to test all the welds of the reactor vessel and its cover of mixed welds of tanks and steam generators, pressurizer welds and gudgeons from the inside; gamma radiography of vessel mixed welds, televisual examination of the stainless steel lining of the reactor vessel and its cover. Use of these technics is made with specific automatic machines designed either for inspection of steam generator tubes or for complete inspection of the vessel. Several reactors were inspected with these devices [fr

Conceptual design of EAST flexible in-vessel inspection system

International Nuclear Information System (INIS)

Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

2010-01-01

Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

Ultrasonic inspection of the Calder Hall and Chaplecross reactor pressure vessels

International Nuclear Information System (INIS)

Pennick, A.M.

1993-01-01

This paper describes the ultrasonic inspection surveys that have recently been carried out on the Calder Hall and Chapelcross Magnox steel reactor pressure vessels. The development of the inspection system, which is based on the Rediman manipulator and uses the Sonomatic Zipscan equipment and Time-of-Flight diffraction techniques is discussed. The inspection results are presented and compared with the original inspection findings and limiting crack sizes. (author)

Guide to the periodic inspection of nuclear reactor steel pressure vessels

International Nuclear Information System (INIS)

1969-01-01

This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

International Nuclear Information System (INIS)

Burns, D.C.

1980-01-01

There is provided for a reactor vessel inspection device a support ring sized to accommodate the circular path defined by three or more guide studs extending upwardly from the vessel. The support ring has at least three movably mounted guide stud bushings which can be positionally adjusted to align each bushing with one of the studs. When engaged, the guide studs and bushings yield a coarse positioning of the inspection device relative to the reactor vessel. Also provided are three support legs which are clamped to the support ring and dimensioned to an appropriate length. Two of the support legs have shoes clamped thereto, configured to rest on an internal circumferential flange within the reactor vessel. The third support leg is provided with a specially adapted shoe configured to engage a locating element, the exact position of which is known, within the vessel to achieve fine positioning of the inspection device relative to the reactor vessel. The support ring is additionally provided with an annular key which runs longitudinally about its outer periphery. Clamping means utilized to secure the guide stud bushings and the support legs to the support ring are provided with keyways to insure automatic self-alignment when fully tightened. (auth)

30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and... METAL AND NONMETAL MINES Compressed Air and Boilers § 56.13015 Inspection of compressed-air receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels...

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

Damage-tolerant design and inspection philosophy for nuclear and other pressure vessels

International Nuclear Information System (INIS)

Adams, N.J.I.

1980-01-01

Statistical analyses of pressure vessel failure rates indicate that, to date, the record is very good. However, the public hazard and environmental consequences of failure in certain industrial processes now give cause for much greater concern. With the exception of an Appendix in ASME III, the current design codes and requirements for new vessels are all based on the assumption that they are free from cracklike defects, but engineers recognize tht such perfect vessels cannot be manufactured. Taking into account failure mechanisms, material properties, pre- and in-service inspection, proof testing, failure statistics and probabilistic methods, views are put forward on how a damage-tolerant design and inspection philosophy may be developed to reduce further the possibility of ''rogue'' vessel failure. 21 refs

Automated ultrasonic shop inspection of reactor pressure vessel forgings

International Nuclear Information System (INIS)

Farley, J.M.; Dikstra, B.J.; Hanstock, D.J.; Pople, C.H.

1986-01-01

Automated ultrasonic shop inspection utilizing a computer-controlled system is being applied to each of the forgings for the reactor pressure vessel of the proposed Sizewell B PWR power station. Procedures which utilize a combination of high sensitivity shear wave pulse echo, 0 degrees and 70 degrees angled longitudinal waves, tandem and through-thickness arrays have been developed to provide comprehensive coverage and an overall reliability of inspection comparable to the best achieved in UKAEA defect detection trials and in PISC II. This paper describes the ultrasonic techniques, the automated system (its design, commissioning and testing), validation and the progress of the inspections

Development of automated ultrasonic device for in-service inspection of ABWR pressure vessel bottom head

International Nuclear Information System (INIS)

Kojima, Y.; Matsuyama, A.

1995-01-01

An automated device and its controller have been developed for the bottom head weld examination of pressure vessel of Advanced Boiling Water Reactor (ABWR). The internal pump casings and the housings of control rod prevent a conventional ultrasonic device from scanning the required inspection zone. With this reason, it is required to develop a new device to examine the bottom head area of ABWR. The developed device is characterized by the following features. (1) Composed of a mother vehicle and a compact inspection vehicle. They are connected only by an electric wire without using the conventional arm mechanism. (2) The mother vehicle travels on a track and lift up the inspection vehicle to the vessel. (3) The mother vehicle can automatically attach the inspection vehicle to the bottom head, and detach the inspection vehicle from it. (4) Collision avoidance control function with a touch sensor is installed at the front of the inspection vehicle. The device was successfully demonstrated using a mock-up of reactor pressure vessel

Analysis and optimization on in-vessel inspection robotic system for EAST

International Nuclear Information System (INIS)

Zhang, Weijun; Zhou, Zeyu; Yuan, Jianjun; Du, Liang; Mao, Ziming

2015-01-01

Since China has successfully built her first Experimental Advanced Superconducting TOKAMAK (EAST) several years ago, great interest and demand have been increasing in robotic in-vessel inspection/operation systems, by which an observation of in-vessel physical phenomenon, collection of visual information, 3D mapping and localization, even maintenance are to be possible. However, it has been raising many challenges to implement a practical and robust robotic system, due to a lot of complex constraints and expectations, e.g., high remanent working temperature (100 °C) and vacuum (10"−"3 pa) environment even in the rest interval between plasma discharge experiments, close-up and precise inspection, operation efficiency, besides a general kinematic requirement of D shape irregular vessel. In this paper we propose an upgraded robotic system with redundant degrees of freedom (DOF) manipulator combined with a binocular vision system at the tip and a virtual reality system. A comprehensive comparison and discussion are given on the necessity and main function of the binocular vision system, path planning for inspection, fast localization, inspection efficiency and success rate in time, optimization of kinematic configuration, and the possibility of underactuated mechanism. A detailed design, implementation, and experiments of the binocular vision system together with the recent development progress of the whole robotic system are reported in the later part of the paper, while, future work and expectation are described in the end.

Analysis and optimization on in-vessel inspection robotic system for EAST

Energy Technology Data Exchange (ETDEWEB)

Zhang, Weijun, E-mail: zhangweijun@sjtu.edu.cn; Zhou, Zeyu; Yuan, Jianjun; Du, Liang; Mao, Ziming

2015-12-15

Since China has successfully built her first Experimental Advanced Superconducting TOKAMAK (EAST) several years ago, great interest and demand have been increasing in robotic in-vessel inspection/operation systems, by which an observation of in-vessel physical phenomenon, collection of visual information, 3D mapping and localization, even maintenance are to be possible. However, it has been raising many challenges to implement a practical and robust robotic system, due to a lot of complex constraints and expectations, e.g., high remanent working temperature (100 °C) and vacuum (10{sup −3} pa) environment even in the rest interval between plasma discharge experiments, close-up and precise inspection, operation efficiency, besides a general kinematic requirement of D shape irregular vessel. In this paper we propose an upgraded robotic system with redundant degrees of freedom (DOF) manipulator combined with a binocular vision system at the tip and a virtual reality system. A comprehensive comparison and discussion are given on the necessity and main function of the binocular vision system, path planning for inspection, fast localization, inspection efficiency and success rate in time, optimization of kinematic configuration, and the possibility of underactuated mechanism. A detailed design, implementation, and experiments of the binocular vision system together with the recent development progress of the whole robotic system are reported in the later part of the paper, while, future work and expectation are described in the end.

78 FR 47716 - Final Guidance Regarding Voluntary Inspection of Vessels for Compliance With the Maritime Labour...

Science.gov (United States)

2013-08-06

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2012-1066] Final Guidance Regarding Voluntary Inspection of Vessels for Compliance With the Maritime Labour Convention, 2006 AGENCY: Coast Guard... procedures regarding the inspection of U.S. vessels for voluntary compliance with the Maritime Labour...

Vessel Sewage Discharges: Statutes, Regulations, and Related Laws and Treaties

Science.gov (United States)

Vessel sewage discharges can be regulated under multiple statutes, regulations, and laws/treaties, including the Clean Water Act, Title XIV, MARPOL Annex IV and the Vessel General Permit. This page describes how these are applied to vessel sewage.

Experience with the WWER-440 MW reactor pressure vessel in-service inspections and evaluation of their results

International Nuclear Information System (INIS)

Brumovsky, M.; Kralovec, J.; Prepechal, J.; Sulc, J.

1989-01-01

The Power Machinery Plant of Skoda Works in Plzen carries out in-service inspections of WWER-440 MW reactor pressure vessels by means of remote controlled inspection equipment - the TRC reactor test system, and some other inspections devices. The results of the in-service inspections were evaluated by methods based on the fracture mechanics approach, the knowledge of stress and strain distribution, and the operating history of the pressure vessels. Examples of types of defects found and their analysis are shown. (author). 1 tab

Application of ultrasonic phased array technique for inspection of stud bolts in nuclear reactor vessel

International Nuclear Information System (INIS)

Choi, Sang Woo; Lee, Joon Ho; Park, Min Su; Cho, Youn Ho; Park, Moon Ho

2004-01-01

The stud bolt is one of crucial parts for safety of reactor vessels in nuclear power plants. Cracks initiation and propagation were reported in stud bolts using closure of reactor vessel and head. Stud bolts are inspected by ultrasonic technique during overhaul periodically for the prevention of stud bolt failure and radioactive leakage from nuclear reactor. In conventional ultrasonic testing for inspection of stud bolts, crack was detected by using shadow effect. It take too much time to inspect stud bolt by using conventional ultrasonic technique. In addition, there were numerous spurious signal reflected from every thread. In this study, the advanced ultrasonic phased array technique was introduced for inspect stud bolts. The phased array technique provide fast inspection and high detectability of defects. There are sector scanning and linear scanning method in phased array technique, and these scanning methods were applied to inspect stud bolt and detectability was investigated.

Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

International Nuclear Information System (INIS)

Sweeney, F.J.; Carroll, D.G.; Chen, C.; Crane, C.; Dalton, R.; Taylor, J.R.; Tosunoglu, S.; Weymouth, T.

1993-01-01

One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS

Apparatus for locating inspection device in a nuclear reactor vessel

International Nuclear Information System (INIS)

1980-01-01

A method for accurately locating an inspection device with a PWR or BWR pressure vessel uses a plurity of guide members and an internal location element, the exact position of which is known. Used for defining the size, orientation and position of a flow. (U.K.)

Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

International Nuclear Information System (INIS)

Saernmark, Ivan; Lenz, Herbert

2008-01-01

Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction

Inspection device for external examination of pressure vessels, preferably for ultrasonic testing of reactor vessels

International Nuclear Information System (INIS)

Figlhuber, D.; Gallwas, J.; Weber, R.; Weber, J.

1978-01-01

The inspection device is placed in the annular gap between pressure vessel and biological shield of the BWR. In the annulus there is arranged at least one longitudinal rail which has got vertical guideways. Along it there can be moved on testing paths a manipulator with the ultrasonic search unit. The manipulator drive is outside of the inspection annulus. It is coupled to the manipulator by means of a tension member being guided over a reversing unit mounted at the upper end of the longitudinal rail. As a tension member there may be used a drag chain; the drive and the reversing unit are provided with corresponding chain wheels. (DG) [de

Medical inspection according to the new nuclear power law

International Nuclear Information System (INIS)

Ligteringen, J.

1982-01-01

Industrial medicine is an underdeveloped discipline in the Netherlands. The author surveys the discrepancies between the laws regulating health protection by medical inspection, reporting and evaluation, and the laws concerning these points when work with radioactive materials or apparatuses producing ionising radiations is involved. Points to be cleared up are presented. (Auth.)

33 CFR 1.01-20 - Officer in Charge, Marine Inspection.

Science.gov (United States)

2010-07-01

... that they comply with the applicable laws, rules, and regulations relating to safe construction... action in cases of misconduct, negligence, or incompetence of merchant marine officers or seamen; and the enforcement of vessel inspection, navigation, and seamen's laws in general. Specific procedures for appealing...

Visual inspection of vessel internals; Visuelle Inspektion von Kerneinbauten

Energy Technology Data Exchange (ETDEWEB)

Rabe, G. [Siemens AG KWU, Erlangen (Germany)

1999-08-01

Visual inspection has matured to a qualified testing method and has become a standard method for inspection of reactor pressure vessels. Until today, all known defects in RPV internals have been detected by visual inspection. The codes KTA 3204 and DIN 25435-4 describe the framework conditions and requirements for visual inspections, which should be adhered to to the most possible extent. Visual inspections are carried by now at all RPV internals, also at those where access is difficult and limited. The inspection robot SUSI is applied in most cases. The camera and manipulator technology meanwhile has been upgraded to a standard performance quality allowing reliable, fast and easy visual inspection. The personnel is trained accordingly, so as to keep abreast with enhancements. Qualification of the inspection system has been simplified and standardised to a large extent. (orig/CB) [Deutsch] Die Sichtpruefung ist zu einem qualifizierten Pruefverfahren gereift und hat bei der Inspektion der RDB-Einbauten einen festen Platz eingenommen. Bisher wurden alle bekannten Schaeden an den RDB-Einbauten bei der Sichtpruefung festgestellt. In der KTA 3204 und der DIN 25435-4 sind die Rahmenbedingungen und Anforderungen an die Sichtpruefung beschrieben, die es gilt, weitestgehend einzuhalten. Mittlerweile werden an allen RDB-Einbauten, auch an den nur bedingt zugaenglichen, Sichtpruefungen vorgenommen. Dabei hat das Inspektionsfahrzeug SUSI inzwischen den breitesten Raum eingenommen. Die Entwicklung der Kamera- und Manipulatortechnik hat inzwischen einen Stand erreicht, der eine sichere, schnelle und einfache Sichtpruefung zulaesst. Das Pruefpersonal wird laufend fuer die Sichtpruefung geschult und qualifiziert. Die Qualifizierung des Inspektionssystems wurde weitestgehend vereinfacht und standardisiert. (orig.)

Guiding device for a manipulator mast for internal inspection of a reactor pressure vessel

International Nuclear Information System (INIS)

Seifert, W.; Schlueter, H.

1977-01-01

A remote-controlled supporting device centering a manipulator mast is described which is mounted and operated above a reactor pressure vessel under water in such a way that rotations and vertical movements necessary for the internal inspection of the pressure vessel remain possible. (RW) [de

30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and...-UNDERGROUND METAL AND NONMETAL MINES Compressed Air and Boilers § 57.13015 Inspection of compressed-air receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure...

Impact of inservice inspection on the reliability of pressure vessels and piping

International Nuclear Information System (INIS)

Bush, S.H.

1975-01-01

The reliability of pressure components of a nuclear reactor is a function of the as-fabricated quality plus a program of continuing inspection during operation. Since insufficient data exist to quantitatively determine failure probabilities of nuclear pressure vessels and piping, it is necessary to utilize information from comparable non-nuclear systems such as power boilers. Based on probabilistic studies it is inferred that in-service inspection improves component reliability one-to-two orders of magnitude depending on the type and completeness of the inspections. An attempt is made to assess the significance of the ASME Section XI Code as to relative completeness of inspection and the probable improvement in reliability. (U.S.)

Impact of inservice inspection on the reliability of pressure vessels and piping

International Nuclear Information System (INIS)

Bush, S.H.

1974-01-01

The reliability of pressure components of a nuclear reactor is a function of the quality as-fabricated plus a program of continuing inspection during operation. Since insufficient data exist to quantitatively determine failure probabilities of nuclear pressure vessels and piping, it is necessary to utilize information from comparable non-nuclear systems such as power boilers. Based on probabilistic studies it is inferred that in-service inspection improves component reliability one-to-two orders of magnitude depending on the type and completeness of the inspections. An attempt is made to assess the significance of the ASME Section XI Code as to relative completeness of inspection and the probable improvement in reliability. (U.S.)

Apparatus for carrying out ultrasonic inspection of pressure vessels

International Nuclear Information System (INIS)

Dent, K.H.; Challender, R.S.

1975-01-01

Apparatus is described for use in carrying out ultrasonic inspection of coolant nozzles of nuclear reactor pressure vessels. It comprises a manipulator for supporting an ultrasonic scanning transducer within the coolant nozzle. The manipulator is carried by a support located within the pressure vessel and comprises a pair of legs pivotable in caliper manner to span the base of the nozzle. Means are provided for pivoting the legs together to enable free entry of the manipulator and scanning transducer into the nozzle, and for pivoting the legs apart to bring the transducer into an operating position adjacent to the wall of the nozzle. The manipulator is rotatable within the nozzle to enable scanning of its interior surface. (U.K.)

Emergency retraction mechanism for the manipulator arm of a nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

1980-01-01

Nuclear reactor vessels are made using numerous welds. These have to be inspected, often using ultrasonic transducers mounted on a manipulator arm. This invention seeks to solve the problem of retracting the manipulator arm should an emergency occur while it is fully extended, particularly within one of the reactor vessel nozzles. Of specific concern is the situation where power fails with the manipulator arm so extended. Details are given of an emergency retraction mechanism for use in reactor vessel inspection apparatus. A manual retraction means is used; the manipulator arm is slidably mounted within a frame. This comprises a member mounted on the arm for looping engagement by a cable, the cable being fixed at one end of the arm frame and engaging the member, and a clamp for detachably securing the cable at its other end to the arm frame at a point which is accessible from above the vessel. (U.K.)

Kinematic and dynamic analysis of a serial-link robot for inspection process in EAST vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Peng Xuebing, E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Shushanhu Road 350, Hefei, Anhui 230031 (China); Yuan Jianjun; Zhang Weijun [Research Institute of Robotics, Mechanical Engineering School, Shanghai Jiao Tong University, No.800, Dong Chuan Road, Min Hang District, Shanghai 200240 (China); Yang Yang; Song Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Shushanhu Road 350, Hefei, Anhui 230031 (China)

2012-08-15

Highlights: Black-Right-Pointing-Pointer A serial-link robot FIVIR is proposed for inspection of EAST PFCs between plasma shots. Black-Right-Pointing-Pointer The FIVIR is a function modular design and has specially designed curvilinear mechanism for axes 4-6. Black-Right-Pointing-Pointer The D-H coordinate systems, forward and inverse kinematic model can be easily established and solved for the FIVIR. Black-Right-Pointing-Pointer The FIVIR can fulfill the required workspace and has a good dynamic performance in the inspection process. - Abstract: The present paper introduces a serial-link robot which is named flexible in-vessel inspection robot (FIVIR) and developed for Experimental Advanced Superconducting Tokamak (EAST). The task of the robot is to carry process tools, such as viewing camera and leakage detector, to inspect the components installed inside of EAST vacuum vessel. The FIVIR can help to understand the physical phenomena which could be happened in the vacuum vessel during plasma operation and could be one part of EAST remote handling system if needed. The FIVIR was designed with the consideration of having easy control and a good mechanics property which drives it resulted in function modular design. The workspace simulation and kinematic analysis are given in this paper. The dynamic behavior of the FIVIR is studied by multi-body system simulation using ADAMS software. The study result shows the FIVIR has ascendant kinematic and dynamic performance and can fulfill the design requirement for inspection process in EAST vacuum vessel.

Kinematic and dynamic analysis of a serial-link robot for inspection process in EAST vacuum vessel

International Nuclear Information System (INIS)

Peng Xuebing; Yuan Jianjun; Zhang Weijun; Yang Yang; Song Yuntao

2012-01-01

Highlights: ► A serial-link robot FIVIR is proposed for inspection of EAST PFCs between plasma shots. ► The FIVIR is a function modular design and has specially designed curvilinear mechanism for axes 4–6. ► The D-H coordinate systems, forward and inverse kinematic model can be easily established and solved for the FIVIR. ► The FIVIR can fulfill the required workspace and has a good dynamic performance in the inspection process. - Abstract: The present paper introduces a serial-link robot which is named flexible in-vessel inspection robot (FIVIR) and developed for Experimental Advanced Superconducting Tokamak (EAST). The task of the robot is to carry process tools, such as viewing camera and leakage detector, to inspect the components installed inside of EAST vacuum vessel. The FIVIR can help to understand the physical phenomena which could be happened in the vacuum vessel during plasma operation and could be one part of EAST remote handling system if needed. The FIVIR was designed with the consideration of having easy control and a good mechanics property which drives it resulted in function modular design. The workspace simulation and kinematic analysis are given in this paper. The dynamic behavior of the FIVIR is studied by multi-body system simulation using ADAMS software. The study result shows the FIVIR has ascendant kinematic and dynamic performance and can fulfill the design requirement for inspection process in EAST vacuum vessel.

Dynamic navigation simulation of an articulated multi-link arm for in-vessel inspection tasks in a Tokamak

International Nuclear Information System (INIS)

Rastogi, Naveen; Prakash, Ravi; Dutta, Pramit; Virpara, Nirav

2013-01-01

This paper is part of the remote handling (RH) activities towards preparedness for the future fusion machines. The aim of the R and D program performed is to demonstrate the feasibility of the inspection tasks inside the Vacuum Vessel. Due to the toroidal geometry and huge dimensions of the Vacuum Vessel, there is an inevitable need of a precise and fast automated articulated inspection system that can perform the required inspection tasks without damaging the surface and to maintain the machine availability for the maximum time. When considering generic Tokamak relevant conditions, the set of major challenges for the Remote Equipment is to sustain the severe operating conditions: ultra high vacuum, temperature and tritium level. The limited number of machine access ports and the very constrained environment complicate the introduction of a robot into the machine. The Multi Link inspection arm is required to be deployed in the bounded environment inside the Tokamak Vessel. This paper presents the development of software for implementation of autonomous navigation motion control algorithms based design simulations for the inspection Arm for routine inspection tasks, navigation to the targeted coordinates inside the Vessel, clash avoidance and to perform other auxiliary mechanical tasks. The software is also capable to store the joints and frame locations at every interval which can be used for the real time control application. The developed software has the flexibility to work with any number of links and joints. (author)

Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

International Nuclear Information System (INIS)

Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I.

1999-12-01

One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (author)

Motion correction for passive radiation imaging of small vessels in ship-to-ship inspections

Energy Technology Data Exchange (ETDEWEB)

Ziock, K.P., E-mail: ziockk@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Boehnen, C.B.; Ernst, J.M.; Fabris, L. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hayward, J.P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Department of Nuclear Engineering, University of Tennessee, Knoxville, TN (United States); Karnowski, T.P.; Paquit, V.C.; Patlolla, D.R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Trombino, D.G. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

2016-01-01

Passive radiation detection remains one of the most acceptable means of ascertaining the presence of illicit nuclear materials. In maritime applications it is most effective against small to moderately sized vessels, where attenuation in the target vessel is of less concern. Unfortunately, imaging methods that can remove source confusion, localize a source, and avoid other systematic detection issues cannot be easily applied in ship-to-ship inspections because relative motion of the vessels blurs the results over many pixels, significantly reducing system sensitivity. This is particularly true for the smaller watercraft, where passive inspections are most valuable. We have developed a combined gamma-ray, stereo visible-light imaging system that addresses this problem. Data from the stereo imager are used to track the relative location and orientation of the target vessel in the field of view of a coded-aperture gamma-ray imager. Using this information, short-exposure gamma-ray images are projected onto the target vessel using simple tomographic back-projection techniques, revealing the location of any sources within the target. The complex autonomous tracking and image reconstruction system runs in real time on a 48-core workstation that deploys with the system.

Development and deployment of BARC Vessel Inspection System (BARVIS) for TAPS-1 and 2

International Nuclear Information System (INIS)

Singh, Jit Pal; Ranjon, R.; Kulkarni, M.P.; Soni, N.L.; Patel, R.J.

2016-01-01

As per regulatory requirements, inspection of welds in Reactor Pressure Vessel (RPV) is necessary for further continuing operation of TAPS - 1 and 2. Upper shell longitudinal welds of RPV which were seen as inaccessible up till now have been inspected first time since operation of the reactors by deploying Weld Inspection Manipulator (WIM) in Unit-1 in August 2012. Subsequently Unit-2 and again Unit-1 upper shell welds were inspected with upgraded versions of WIM in Feb 2013 and March 2015 respectively. Inspection of upper shell welds paves the way for more challenging inspection of beltline region welds. These welds are accessible only from Inside Diameter (JD) surface through a narrow annular gap of 25 mm between RPV wall and thermal shield by managing obstructions due to core internals. BARC Vessel Inspection System (BARVIS) for inspection of beltline region welds from inner side of the RPV was designed, manufactured, tested and qualified for sending scanning probes in to the annular gap of 25mm as per RPV engineering drawings and also based on actual gap measured in Unit-2. Annular gap measurement was done in Units-1 before deployment of BARVIS during 25 th Refuelling Outage (RFO) in Unit-1. Contrary to the expectation, the annular gap was found less and inspection of beltline region with BARVIS in Unit-1 could not be done. Finally during RFO in January 2016 of Unit-2, BARVIS has been successfully deployed for beltline region welds inspection. BARVIS mainly consists of manipulator, its operating system and data acquisition system. Data acquisition system and data analysis are not covered in this report. (author)

Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

1999-12-01

One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

Wine vessels (Vasa vinaria in roman law

Directory of Open Access Journals (Sweden)

Aličić Samir

2017-01-01

Full Text Available The notion of 'wine vessels' in Roman law comprises all the winecontaining recipients. There is no legal standardization of wine vessels by means of volume, and although the terms amphora, urna and culleus are used to designate both the vessels and the units of measure, these are two different meanings of the terms. In regard of the question, whether the vessels make appurtenance of the wine, jurisprudents of proculean school divided them in two categories. In the first category are those that follow legal status of wine, usually amphoras and other jars (cadi which are used for 'packaging', i. e. 'bottling' of the wine. The second category make mostly vats (cuppae and ceramic cisterns (dolia, which don't follow legal status of wine, making instead part of farming equipment of a landed property (instrumentum fundi and it's appurtenance. But, the roman jurists are not consistent regarding criteria for distinguishing these two categories.

78 FR 13691 - Guidance Regarding Inspection and Certification of Vessels Under the Maritime Security Program

Science.gov (United States)

2013-02-28

...-level approach would enable the Coast Guard to apply traditional inspection methods to newly reflagged... enrolled MSP vessels would continue to be inspected by the Coast Guard in a manner similar to traditional... oversight, and evaluations of Authorized Class Society (ACS) survey activities. We received eight public...

To the application of TV and optical equipment for in-service inspection of reactor vessel and primary circuit component materials

International Nuclear Information System (INIS)

Afonin, Eh.M.; Bachelis, I.M.; Tokarev, E.A.; Yastrebov, V.E.

1985-01-01

Some problems of application of TV and optical equipment for inspection of reactor vessel and primary circuit component materials are considered taking the most widespread WWER-440 type reactor as an example. The most advanrageous objects of the inspection and typical zones of equipment arrangement are shown. Methods and peculiarities of the inspection with the use of TV and optical equipment are presented. Recommendations on rational application of the equipment for the inspection of WWER-440 reactor vessel components are given

Opinions of the well-informed persons about the nuclear reactor facility periodical inspection

International Nuclear Information System (INIS)

Aeba, Yoichi; Ishikawa, Michio; Enomoto, Toshiaki; Oomori, Katsuyoshi

2005-01-01

Falsifications of self-inspection records in the shrouds and of leakage rates for containment vessels at TEPCO nuclear power plants destroyed public trust in nuclear safety. The Nuclear Reactor Regulation Law and Electric Utility Law were amended to enhance the nuclear safety regulation system. The major improvements are that operators are legally required to conduct inspection (periodical operator inspection) and recording and keeping inspection results. The operator performs 'periodical operator inspection' regularly, and Nuclear and Industrial Safety Agency (NISA) performs periodical inspection' about particularly important facilities/function in safety. Sixteen opinions of well-informed persons about the nuclear reactor facility periodical inspection were presented in this special number. Interval of periodical inspection less than thirteen months was disputed. Maintenance activities should be more rationalized based on risk information. (T. Tanaka)

Influence of the inservice inspection on the reliability of a pressure vessel under stress corrosion cracking

International Nuclear Information System (INIS)

Francisco, Alexandre S.; Melo, Paulo F.F.

1997-01-01

The influence of in-service inspection on the structural integrity of a pressure vessel subjected to an aggressive environment is assessed in a quantitative manner with the purpose of establishing an optimal inspection program. For illustration purposes, the specific case of a LPG storage vessel under stress corrosion cracking is analysed, for which some failure data is available. Its failure probability is evaluated by means of a model published in the literature, for which three probabilities are taken into account: the first, concerns the distribution of initial crack depths; the second one accounts for the probability of detecting these cracks by nondestructive testing and finally, the probability of not detecting a crack until the next inspection is undertaken. The analysis is performed by means of a trade-off between two important parameters: the inspection sensitivity and the inspection interval. The influence of these parameters on the probability of failure is displayed in a graphical fashion where the benefits which can be achieved in various inspections programs are clearly revealed. Decisions can then be taken as to the optimal parameters. (author). 8 refs., 2 figs

Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

Energy Technology Data Exchange (ETDEWEB)

Saernmark, Ivan; Lenz, Herbert [WesDyne TRC AB, Stockholm (Sweden)

2008-04-15

Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction.

46 CFR 167.15-15 - Application for inspection of a new nautical school ship or a conversion of a vessel to a...

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Application for inspection of a new nautical school ship or a conversion of a vessel to a nautical school ship. 167.15-15 Section 167.15-15 Shipping COAST... Inspections § 167.15-15 Application for inspection of a new nautical school ship or a conversion of a vessel...

Experiences concerning reactor pressure vessel head penetration inspections; Erfahrungen mit Pruefungen von Reaktordruckbehaelter-Deckeldurchfuehrungen

Energy Technology Data Exchange (ETDEWEB)

Debnar, Angelika [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2009-07-01

Globally observed damage at the control rod drive mechanism nozzles in PWR-type reactors (Bugey-3, Oconee 1,2,3 and ANO-1, David Besse) have triggered enhanced inspection of reactor pressure vessel (RPV) head penetrations. In Germany the regulations require a periodic inspection especially of dissimilar welds. Westinghouse has developed an automated measuring system for RPV heads aimed to inspect welded joints at open nozzles of nozzles with thermosleeves. The testing technology with remote controlled robotics is supposed to perform a weld inspection as complete as possible, restraints result from constructive difficulties for the accessibility. The new gap-scanner DE2008 was qualified at the mock-up and was implemented into the periodic in-service inspection of Neckarwestheim-1.

The 1978 first in-service inspection of the reactor pressure vessel of the second unit of the Greifswald nuclear power plant

International Nuclear Information System (INIS)

Pastor, D.; Busch, R.; Hildebrandt, E.; Redlich, K.H.

1979-01-01

The reactor pressure vessel and the primary coolant circuit of the second 440-MW(e) unit of the Greifswald nuclear power plant were subjected to an in-service inspection. Extent of the inspection, development and construction of a reactor inspection container as well as the nondestructive materials testing methods used are described. Further, problems of performing the inspection, such as needs of time and personnel and radiation exposure, are considered. Finally, it is stated that the reactor pressure vessel was in safe operating state. (author)

Positioning calibration apparatus for transducers employed in nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Elsner, H.J.

1979-01-01

Calibration apparatus for verifying the position and orientation of transducers used in nuclear reactor vessel inspection apparatus is disclosed. A tank, filled with water, the operating inspection medium, is fitted with a movable mounting assembly adapted to securely accommodate a transducer and the mounting assembly in which it is normally secured during an inspection procedure. The tank is also provided with a slidably mounted target positioned therein at a predetermined distance from the target which is selected to avoid the distortion effects in the near field of the transducer response. The calibration apparatus can be used to check the normal transducer mounting for either perpendicularity or angular orientation by moving the tank's mounting assembly via a lead screw with which it is threadingly engaged. 6 claims

Review of current practices and requirements for the inspection of prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Reimann, K.J.

1980-12-01

Code requirements for pre- and in-service inspection of prestressed concrete pressure vessels as utilized in gas-cooled reactors are reviewed and compared with practices and experiences during construction, commissioning, and operation of such reactors. The pre-service inspection relies heavily on embedded instrumentation for measurements of stresses, temperatures, and displacements. The same instrumentation is later used for in-service surveillance, which additionally includes visual examination of exposed surfaces, monitoring of tendon conditions, and measurement of tendon loads. Improvement of present monitoring instrumentation and/or techniques, rather than development of new in-service inspection methods, is recommended

In service inspection for Superphenix vessels development of ultrasonic techniques available at high temperature

International Nuclear Information System (INIS)

Gondard, C.

1983-12-01

The main and safety vessels of SUPERPHENIX 1 were designed to allow in-service inspections. The remote controlled inspection device MIR was developped for this purpose. The ultrasonic examination has required the development of all new transducers fitted with severe operating conditions prevailing in intervessels interval. A list of problems to be resolved and technological solutions which were found is given. Measurements of acoustical properties on actual probes are compared with theoretical values. It appears that concordance is good and that an in-service inspection using high temperature transducers is possible with a good spatial resolution and signal to noise ratio

In-service inspection of the vessels of nuclear submarine reactors

International Nuclear Information System (INIS)

Saglio, R.; Mercier L'Abbe, G.

1977-02-01

Description is given of a machine of automatic inspection of the reactor vessels of nuclear submarines, that was developed by the French Atomic Energy Commission (CEA). A focused ultrasound testing method is used. The equipment such described meets the safety requirements prescribed for the primary circuits of nuclear submarine boiler plants. The sensitivity and resolution power of the apparatus allow such high precision to be obtained as rupture mechanics gains a credibility that could not have been obtained with using conventional apparatus [fr

Positioning calibration apparatus for transducers employed in nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Elsner, H.J.

1981-01-01

The invention provides a calibration apparatus suitable for verifying the position and orientation of transducers used in reactor vessel ultrasonic inspection. The apparatus includes moveable mounting means which secures a transducer within the tank in its normal inspection orientation. A drive is also provided for moving the transducer in the tank relative to a target. The target is slidably positioned in the tank at a distance from the transducer which is selected to avoid the distortion effects in the near field of the transducer. The drive mechanism may be provided with graduated indicia of travel, or a scale may be affixed to the side of the tank. (L.L.)

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

In-vessel inspection before head removal: TMI II: Phase I. (Conceptual development)

International Nuclear Information System (INIS)

Calloway, N.E.; Greenlee, D.W.; Lawrence, G.R.; Paglia, A.L.; Piatt, T.D.; Tucker, B.A.

1981-08-01

The objective of the task is to provide for an internal inspection of the reactor vessel and the fuel assemblies prior to reactor vessel head removal. Because the degree of damage to equipment and fuel in the TMI-II reactor is not precisely known, it is important that as much information as possible be obtained on present conditions inside the reactor. This information will serve to benchmark the various analyses already completed or underway and will also guide the development of programs to obtain more information on the TMI-II core damage. In addition, the early look will provide data for planning the reactor disassembly program

Cost regulation on the inspection of plants requiring supervision

International Nuclear Information System (INIS)

Anon.

1981-01-01

According to annexes I to VI of this regulation, TUeVs (technical control authorities) (2nd sentence of para. 1 of sect. 24 c of the trade law) collect fees for inspections ordered by the authorities for the following plants and installations: 1. steam boiler plants, 2. pressure vessels, high-pressure gas vessels, feeders, 3. lifts, 4. acetylene plants, 5. plants for the storage, racking and transport of combustile liquids, 6. electrical installations on hazardous location. (orig.) [de

Development of Reactor Vessel Bottom Mount Instrumentation Nozzle Routine Inspection Device

Energy Technology Data Exchange (ETDEWEB)

Khaled, Atya Ahmed Abdallah; Ihn, Namgung [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

The primary coolant water of pressurized water reactors has created cracks in j-weld of penetrations with Alloy 600 through a process called primary water stress corrosion cracking. On October 6, 2013, BMI nozzle number 3 at Palo Verde Unit 3 (PVNGS-3) exhibited small white de-posits around the annulus. Nozzle attachment to the RV lower head is by J-groove weld to the inside penetration of the nozzle and the weld material is of Alloy 600 material. Above two cases clearly show the necessity of routine inspection of RV lower head penetration during refueling outage. Nondestructive inspection is generally performed to detect fine cracks or defects that may develop during operation. Defects usually occur at weld regions, hence most non-destructive inspection is to scan and check any defects or crack in the weld region. BMI nozzles at the bottom head of a nuclear reactor vessel (RV) are one of such area for inspection. But BMI nozzles have not been inspected during regular refuel outage due to the relative small size of BMI nozzle and limited impact of the consequences of BMI leak. However, there is growing concern since there have been leaks at nuclear power plants (NPPs) as well as recent operating experience. In this study, we propose a system that is conveniently used for nondestructive inspection of BMI nozzles during regular refueling outage without removing all the reactor internals. A 3D model of the inspection system was also developed along with the RV and internals which permits a virtual 3D simulation to check the design concept and usability of the system.

Research and development on in-service inspection system for reactor vessel of FBR's

International Nuclear Information System (INIS)

Rindo, Hiroshi; Mitabe, Noriaki; Ara, Kuniaki; Nagai, Keiichi; Otaka, Masahiko

1993-01-01

In-Service Inspection (ISI) is required for main components and piping of FBRs. Visual test and volumetric examination of the reactor vessel (RV) from the outer surface are to be performed under severe conditions such as limited space, high temperature and high gamma dose rate during the reactor shutdown. Therefore, ISI should be performed by using a remote operation system, and the ISI system should be very compact. PNC has been developing the ISI system to apply to the RV inspection. Verification and performance tests of ISI system were carried out by use of the RV test model. This paper describes the system structure, system verification tests including operation and controlling the inspection robot, the functions of the visual test and the volumetric examination under the high temperature

Operational aspects of the Calder Hall and Chapelcross pressure vessel ultrasonic inspections

International Nuclear Information System (INIS)

Bithell, S.J.; Howard, S.R.

1993-01-01

As a consequence of the NII's assessment of the Calder Hall and Chapelcross Long Term Safety Review, BNFplc were required to demonstrate the integrity of the Reactor Pressure Vessels through a programme of volumetric seam weld inspection. Existing equipment proved to be inadequate and necessitated the design and manufacture of a remote power manipulator and ultrasonic scanning package. Calder Hall Operations Department and Sellafield Technical Department, working closely with contract staff, completed the first stage of this technically demanding task within 14 months of the project's initiation, resulting in the first deployment of ''REDIMAN'' in March 1991. The design of the new equipment, and the technical and operational difficulties which were overcome by the Inspection Team are outlined. (author)

76 FR 75948 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel CHRYSALIS; Invitation for...

Science.gov (United States)

2011-12-05

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2011 0149] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel CHRYSALIS; Invitation for Public Comments AGENCY... CHRYSALIS is: Intended Commercial Use of Vessel: ``Weekly charter vessel.'' Geographic Region: ``Florida...

Incore inspection device

International Nuclear Information System (INIS)

Ogisu, Tatsuki; Taguchi, Kosei.

1995-01-01

The device of the present invention can inspect surfaces of equipments in reactor water in a nuclear reactor in a state of atmospheric air. Namely, an inspection device is movable forwardly and backwardly in a water-proof vessel. An annular sucker with pleats is disposed to the outer side of a lid of the water-proof vessel. A television camera for an under water monitoring is disposed to the inner side of the lid of the water-proof vessel by way of a partitioning wall with lid. Transferring screws are disposed at the back and on the side of the water-proof vessel. In the device having such a constitution, (1) the inside of the water-proof vessel is at first made water-tight by closing the partitioning wall with lid, (2) the back and the side screws are operated by the guide of the underwater monitoring television camera, to transfer the water-proof vessel to the surface of the reactor core to be inspected, (3) the annular sucker with pleats is urged on the surface to be inspected by the back screw, to fix the water-proof vessel, (4) reactor water in a space of the annular sucker with pleats is discharged and replaced with air, and (5) the lid of the partition wall with lid is opened and the inspection device is disposed at a position of the underwater monitoring television camera, to inspect the surface to be inspected in a state of atmospheric air. (I.S.)

78 FR 46410 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel IMPOSSIBLE DREAM; Invitation...

Science.gov (United States)

2013-07-31

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0084] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel IMPOSSIBLE DREAM; Invitation for Public Comments AGENCY... of the vessel IMPOSSIBLE DREAM is: Intended Commercial Use of Vessel: ``Day, Multiple Day, Week...

Design and implementation of motion planning of inspection and maintenance robot for ITER-like vessel

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng; Lai, Yinping [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai 200240 (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China (China); Cao, Qixin [Institute of Robotics, Shanghai Jiao Tong University, Shanghai 200240 (China)

2015-12-15

Robot motion planning is a fundamental problem to ensure the robot executing the task without clashes, fast and accurately in a special environment. In this paper, a motion planning of a 12 DOFs remote handling robot used for inspecting the working state of the ITER-like vessel and maintaining key device components is proposed and implemented. Firstly, the forward and inverse kinematics are given by analytic method. The work space and posture space of this manipulator are both considered. Then the motion planning is divided into three stages: coming out of the cassette mover, moving along the in-vessel center line, and inspecting the D-shape section. Lastly, the result of experiments verified the performance of the motion design method. In addition, the task of unscrewing/screwing the screw demonstrated the feasibility of system in function.

Ultrasonic Inspection of Cracks in Stud Bolts of Reactor Vessels in Nuclear Power Plants by Signal Processing of Differential Operation

International Nuclear Information System (INIS)

Choi, Sang Woo; Lee, Joon Hyun; Oh, Won Deok

2005-01-01

The stud bolt is one of crucial parts for safe operation of reactor vessels in nuclear power plants, Crack initiation and propagation were reported in stud bolts that arc used for closure of reactor vessel and head, Stud bolts are inspected by ultrasonic technique during overhaul periodically for the prevention of stud bolt failure which could induce radioactive leakage from nuclear reactor, In conventional ultrasonic testing for inspection of stud bolts, cracks are detected by using shadow effect It takes too much time to inspect stud bolts by using conventional ultrasonic technique. In addition, there were numerous spurious signals reflected from every oblique surfaces of thread, In this study, the signal processing technique for enhancing conventional ultrasonic technique was introduced for inspecting stud bolts. The signal processing technique provides removing spurious signal reflected from every oblique surfaces of thread and enhances detectability of defects. Detectability for small crack was enhanced by using this signal processing in ultrasonic inspection of stud bolts in Nuclear Power Plants

The influence of frequency and reliability of in-service inspection on reactor pressure vessel disruptive failure probability

International Nuclear Information System (INIS)

Jordan, G.M.

1978-01-01

A simple probabilistic methodology is used to investigate the benefit, in terms of reduction of disruptive failure probability, which comes from the application of periodic in-service inspection to nuclear pressure vessels. The analysis indicates the strong interaction between inspection benefit and the intrinsic quality of the structure. In order to quantify the inspection benefit, assumptions are made which allow the quality to be characterised in terms of the parameters governing a log normal distribution of time - to - failure. Using these assumptions, it is shown that the overall benefit of in-service inspection unlikely to exceed an order of magnitude in terms of reduction of disruptive failure probability. The method is extended to evaluate the effect of the periodicity and reliability of the inspection process itself. (author)

Remote handling and robotic inspections of Palo Verde reactor vessel internals

International Nuclear Information System (INIS)

Ryder, W.

1998-01-01

Remote visual examinations and handling evolutions in high radiation field environments have required the use of radiation tolerant video systems. These systems involve significant expense and potentially require large envelope deployment structures. Recent events at Palo Verde including Upper Guide Structure damage and Reactor Vessel In-Service Inspections have provided opportunities for research, design and utilization of alternative approaches. Most significant of these, utilization of CCD modules with high magnification capabilities, have produced higher quality viewing, reduced maintenance expenditures, and rapid deployment intervals. (orig.) [de

Inspecting nuclear pressure vessels: the conundrum of minimizing risk

International Nuclear Information System (INIS)

Oestberg, G.

1992-01-01

The probability of a sudden, massive release of radioactivity from a light-water nuclear reactor through a breach of the containment is assessed on the basis of statistical data which partly consist of subjective estimates. This breach refers to the existence of crack-like defects remaining after a non-destructive examination of the main pressure vessel surrounding the reactor core. Two studies have recently been made of such sources of information about the effectiveness of non-destructive examination of pressure vessels with respect to defects. The results of these studies indicate that the data used as input in the probabilistic calculations do not possess the reliability that might be assumed from the assessments. This type of failure should therefore no longer be considered a de minimis case. In the present review the overconfidence in the efficiency of non-destructive examination is discussed from psychological, sociological and political science points of view. It is concluded that ingrained professional assumptions and values seem to be the main reason for the trust in the technology of inspection. However, there are also psychological constraints that can be understood only in their social and political contexts. (author)

Development of an automated remote inspection system for the interior of the primary containment vessel of a nuclear power plant

International Nuclear Information System (INIS)

Senoo, Makoto; Yoshida, Tomiharu; Omote, Tatsuyuki; Tanaka, Keiji; Koga, Kazunori

1996-01-01

An automated remote inspection system has been developed for the interior of the primary containment vessel of a nuclear power plant. This system consists of an inspection robot and an operator's console. The inspection robot travels along a monorail provided in the interior of the primary containment vessel. The operator's console is located in the central control room of the power plant. We have made efforts to downsize the robot and automate the inspection and monitoring machinery. As for downsizing the robot, a 152 mm wide, 290 mm high cross-sectional area and 15 kg weight can be realized using commercially available small sensors and rearranging the parts in those sensors. As for automating the inspection and monitoring, several monitoring functions are developed using image processing, frequency analysis and other techniques applied to signals from sensors such as an ITV camera, an infrared camera and a microphone, which are mounted on the robot. Endurance tests show resistance of the robot to radiational and thermal conditions is adequate for actual use in actual power plants. (author)

76 FR 67255 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel KIWI CAT; Invitation for...

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2011-10-31

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2011 0135] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel KIWI CAT; Invitation for Public Comments AGENCY... CAT is: INTENDED COMMERCIAL USE OF VESSEL: ``Small vessel 12 person chartering service for the San...

77 FR 5628 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BIG GAME; Invitation for...

Science.gov (United States)

2012-02-03

... Administrative Waiver of the Coastwise Trade Laws: Vessel BIG GAME; Invitation for Public Comments AGENCY....gov . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel BIG GAME is: Intended Commercial Use of Vessel: ``Fishing charter boat.'' Geographic Region: ``Rhode Island...

In-service inspection as an aid to steel pressure vessel reliability

International Nuclear Information System (INIS)

Nichols, R.W.

1975-01-01

In-service inspection has played an important role in non-nuclear pressure vessel technology, being a legal requirement in many countries. Evidence from surveys of reliability of non-nuclear plant has suggested that such inspections can be effective in reducing the risk of subsequent failures. Recent requirements of the ASME XI code which will be summarised have important implications on the techniques to be used for in-service inspection, and so on design and fabrication aspects. Moreover, in-service inspection can only be an effective procedure if its possible weaknesses are recognised. The first problem is to ensure that an ultrasonic technique is used which is capable of detecting defects of an order of magnitude smaller than the critical size for each particular situation, in whatever defect orientation is important. The potential of different ultrasonic techniques will be compared. Next it is necessary to ensure coverage of all the relevant material. In this respect machine operation is superior to manual scanning, so that manipulation and scanning devices have to be developed. Problems of local geometry and of deviations in geometry have to be discussed with designer and fabricator; plate and clad quality have to be controlled (with respect to surface contour, metallurgical condition and freedom from interfering defects) to ensure inspectability in depth. The reliability of the mechanical and electronic equipment has to be assessed and designed to meet high requirements. Some presentational aids to detection and interpretation will be discussed. Having located a potential defect, the application of fracture mechanics treatments requires knowledge of size, shape and orientation. Some of the problems will be discussed together with possible solutions. (author)

Conceptual design of an in-vessel inspection robotic system for Tokamak environment

International Nuclear Information System (INIS)

Kumar, Prabhat; Raju, Daniel; Ranjan, Vaibhav; Patel, Prateek; Dave, Jatinkumar; Naik, Mehul

2013-01-01

An in-vessel inspection robotic system has been conceptualized for operation inside a tokamak vessel. The robotic system is envisaged to comprise of a robotic arm, end-effector, microcontroller and wireless communication system. The end-effector is envisaged to be a special purpose camera for in-situ inspection between plasma shots. The three-link robotic arm, designed for ITER-like environment, has 4 revolute joints- 3 providing manipulation in poloidal plane and the fourth one providing limited movement in adjacent toroidal planes. This paper provides the conceptual design of the system along with kinematic analysis of robotic arm. Solutions have been derived for forward and inverse kinematic models and the Jacobian matrix for the robotic arm linkage. In forward kinematic model, given a set of joint-link parameters, the position and orientation of end-effector are determined with respect to a reference frame. In inverse kinematic model, given the specified position and orientation of end-effector with respect to a reference frame, a set of joint variables are derived that would bring the end-effector into the required posture. Using Jacobian matrix, the relation between the end-effector velocity and the joint velocity of a manipulator is obtained i.e. given the individual joint velocity; the end-effector velocity is obtained. A CAD model has been generated using CATIA to simulate the kinematic model and carry out computational stress analysis. (author)

77 FR 31432 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel PASSION; Invitation for...

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2012-05-25

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2012 0062] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel PASSION; Invitation for Public Comments AGENCY... PASSION is: Intended Commercial Use of Vessel: ``Bareboat chartering, sailing classes.'' Geographic Region...

78 FR 39061 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel MISTRESS MALLIKA; Invitation...

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2013-06-28

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0076] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel MISTRESS MALLIKA; Invitation for Public Comments AGENCY... MISTRESS MALLIKA is: Intended Commercial Use Of Vessel: ``Pleasure Charters''. Geographic Region: Rhode...

In-service inspection of ET-RR-1 reactor vessels and spent fuel storage tank

International Nuclear Information System (INIS)

Khattab, M.; Shafy, M.; Konoplev, K.; Samodurve, YU.; Orlov, S.; Didenko, V.; Jackorev, O.

1993-01-01

Technical survey included in-service inspection are needed in order to investigate the structural integrity and to insure safe operation of the ET-R R-1 reactor after thirty years aging. An intensive work for the inspection of the inspection of the central tank, shield tank, horizontal channels, primary coolant circuit and spent fuel storage tank have been carried out. The inspection procedures were visual method using video camera and magnification optical as well as thickness measurements using ultrasonic gauge meter and replica for determining defect depth. Water chemical analysis of the primary cooling circuit and spent fuel storage were helpful in results explanation. The results showed that the reactor vessels have good surface conditions. The observed pitting did not affect the structural integrity. The majority of the defects were pits having maximum surface area of about 50 mm. Their depth does not exceed 2 mm. The pits depth rate penetration is of the order of 0.5% per year. Thickness measurements showed insignificant variation. Water status and its chemical properties are very important in controlling corrosion rate. 18 figs., 14 tabs

78 FR 53004 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SECOND WIND; Invitation for...

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2013-08-27

... Administrative Waiver of the Coastwise Trade Laws: Vessel SECOND WIND; Invitation for Public Comments AGENCY... SECOND WIND is: Intended Commercial Use of Vessel: ``Breakfast, lunch, and dinner cruises featuring... to the docket number of this notice and the vessel name in order for MARAD to properly consider the...

78 FR 40265 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel HMS-DISCOVERY; Invitation for...

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2013-07-03

... Administrative Waiver of the Coastwise Trade Laws: Vessel HMS-DISCOVERY; Invitation for Public Comments AGENCY... of the vessel HMS-DISCOVERY is: Intended Commercial Use of Vessel: ``Sight Seeing Tours''. Geographic... regulations at 46 CFR Part 388. Privacy Act Anyone is able to search the electronic form of all comments...

The influence of frequency and reliability of in-service inspection on reactor pressure vessel disruptive failure probability

International Nuclear Information System (INIS)

Jordan, G.M.

1977-01-01

A simple probabilistic methodology is used to investigate the benefit, in terms of reduction of disruptive failure probability, which comes from the application of periodic In Service Inspection (ISI) to nuclear pressure vessels. The analysis indicates the strong interaction between inspection benefit and the intrinsic quality of the structure. In order to quantify the inspection benefit, assumptions are made which allow the quality to be characterized in terms of the parameters governing a Log Normal distribution of time-to-failure. Using these assumptions, it is shown that the overall benefit of ISI is unlikely to exceed an order of magnitude in terms of reduction of disruptive failure probability. The method is extended to evaluate the effect of the periodicity and reliability of the inspection process itself. (author)

Development of a Remotely-operated Visual Inspection System for Reactor Vessel Bottommounted Instrument Penetrations of KSNP and Lessons Learned

International Nuclear Information System (INIS)

Jeong, Kyungmin; Choi, Youngsu; Lee, Sunguk; Seo, Yongchil; Kang, Jong Gyu; Kim, Seungho; Jung, Seungho

2006-01-01

In April 2003, South Texas Project Unit 1 made a surprising discovery of boron acid leakage from two nozzles from a bare-metal examination of the reactor vessel bottom-mounted instrument penetrations during a routine refueling outage. A small powdery substance about 150mg was found on the outside of two instrument guide penetration nozzles on the bottom of the reactor. The primary coolant water of pressurized water reactors has caused cracking in penetrations with Alloy 600 through a process called primary water stress corrosion cracking. In South Korea, it is required to conduct 100% visual inspection of the outside of instrument guide penetration nozzles on the bottom of PWRs to confirm the integrity of reactor vessel. This paper describes the remotely-operated visual inspection systems for reactor vessel bottom-mounted instrument penetrations dispatched two times to Youngkwang NPPs and discusses the lessons learned

78 FR 9709 - Draft Guidance Regarding Voluntary Inspection of Vessels for Compliance With the Maritime Labour...

Science.gov (United States)

2013-02-11

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2012-1066] Draft Guidance Regarding Voluntary Inspection of Vessels for Compliance With the Maritime Labour Convention, 2006 AGENCY: Coast Guard... Labour Convention, 2006 (Convention), which enters into force on August 20, 2013. The NVIC will provide...

Improvements in the UT Inspection of vessel nozzles. Array application

International Nuclear Information System (INIS)

Tanarro, A.; Garcia, A.; Izquierdo, J.

1998-01-01

Automatic ultrasonic inspection of certain components in nuclear power plants, together with problems related to access of same, result in other difficulties due to the complexity of their geometry and the apparent orientation of possible defects. Array technology, recently developed on the basis of the theoretical principals of phased array technique, has meant that it is now possible to advance in the characterisation, localisation, and sizing of the defects in these components. This has been possible thanks to the discovery of synthetic materials which have allowed us to design and manufacture a new group of ultrasonic transducers. To these we may add new developments in electronics and computer sciences which have facilitated the building of high-powered control systems. This report discusses the work carried out by Tecnatom and Iberdrola in the field of automatic ultrasonic inspection of the vessel nozzles by means of array technology in the BWR at the Cofrentes Nuclear Power Station. The aims of this work were: - To facilitate the detection, characterisation, sizing and positioning of defects - To simplify and improve ultrasonic inspection in order to reduce acquisition times and the cost of same In order to achieve these results the following items were developed: - New array transducers were designed and manufactured - A new data acquisition system was developed - New programs for analysing data and for simulating ultrasonic testing was developed - The results have been validated in mock up. (Author)

A scaling law for the local CHF on the external bottom side of a fully submerged reactor vessel

International Nuclear Information System (INIS)

Cheung, F.B.; Haddad, K.H.; Liu, Y.C.

1997-01-01

A scaling law for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water has been developed from the results of an advanced hydrodynamic CHF model for pool boiling on a downward facing curved heating surface. The scaling law accounts for the effects of the size of the vessel, the level of liquid subcooling, the intrinsic properties of the fluid, and the spatial variation of the local critical heat flux along the heating surface. It is found that for vessels with diameters considerably larger than the characteristic size of the vapor masses, the size effect on the local critical heat flux is limited almost entirely to the effect of subcooling associated with the local liquid head. When the subcooling effect is accounted for separately, the local CHF limit is nearly independent of the vessel size. Based upon the scaling law developed in this work, it is possible to merge, within the experimental uncertainties, all the available local CHF data obtained for various vessel sizes under both saturated and subcooled boiling conditions into a single curve. Applications of the scaling law to commercial-size vessels have been made for various system pressures and water levels above the heated vessel. Over the range of conditions explored in this study, the local CHF limit is found to increase by a factor of two or more from the bottom center to the upper edge of the vessel. Meanwhile, the critical heat flux at a given angular position of the heated vessel is also found to increase appreciably with the system pressure and the water level

78 FR 14413 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel LUCKY DUCK; Invitation for...

Science.gov (United States)

2013-03-05

... Administrative Waiver of the Coastwise Trade Laws: Vessel LUCKY DUCK; Invitation for Public Comments AGENCY... should refer to docket number MARAD-2013-0019. Written comments may be submitted by hand or by mail to... LUCKY DUCK is: Intended Commercial Use Of Vessel: ``The vessel is to be operated as a sailing...

76 FR 75949 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel PANGAEA; Invitation for...

Science.gov (United States)

2011-12-05

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2011 0150] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel PANGAEA; Invitation for Public Comments AGENCY... PANGAEA is: Intended Commercial Use of Vessel: ``We would like to offer Pangaea for day [[Page 75950...

Flaw distribution development from vessel ISI data

International Nuclear Information System (INIS)

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

1991-01-01

Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed

Qualification of non-destructive examination for belgian nuclear reactor pressure vessel inspection

International Nuclear Information System (INIS)

Couplet, D.; Francoise, T.

2001-01-01

In Service Inspection (ISI) participates to the assessment of Nuclear Reactor Pressure Vessel Integrity. The performance of Non Destructive Examination (NDE) techniques must be demonstrated according to predefined objectives. The qualification process is essential to trust the reliability of the information provided by NDE. In Belgian Nuclear Power Plants, the qualification was conducted through a collaboration between the vendor and a technical group from the Electricity Utility. The important facts of this qualification will be presented: - the detailed definition of the inspection and qualifications objectives, based on a combination of the ASME code and the European Methodology for Qualification; - the systematic verification of the NDE performance and limitations, for each ISI objective, through an adequate combination of tests on blocks and technical justification; - the continuous improvement of the NDE procedure; - the feedback and the lessons learnt from site experience; - the necessary multi-disciplinary approach (NDE, degradation mechanisms, structural integrity)

Vision-Based Corrosion Detection Assisted by a Micro-Aerial Vehicle in a Vessel Inspection Application

Directory of Open Access Journals (Sweden)

Alberto Ortiz

2016-12-01

Full Text Available Vessel maintenance requires periodic visual inspection of the hull in order to detect typical defective situations of steel structures such as, among others, coating breakdown and corrosion. These inspections are typically performed by well-trained surveyors at great cost because of the need for providing access means (e.g., scaffolding and/or cherry pickers that allow the inspector to be at arm’s reach from the structure under inspection. This paper describes a defect detection approach comprising a micro-aerial vehicle which is used to collect images from the surfaces under inspection, particularly focusing on remote areas where the surveyor has no visual access, and a coating breakdown/corrosion detector based on a three-layer feed-forward artificial neural network. As it is discussed in the paper, the success of the inspection process depends not only on the defect detection software but also on a number of assistance functions provided by the control architecture of the aerial platform, whose aim is to improve picture quality. Both aspects of the work are described along the different sections of the paper, as well as the classification performance attained.

76 FR 79764 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel DREAM CATCHER; Invitation for...

Science.gov (United States)

2011-12-22

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2011-0156] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel DREAM CATCHER; Invitation for Public Comments AGENCY... DREAM CATCHER is: INTENDED COMMERCIAL USE OF VESSEL: ``Passenger charter.'' GEOGRAPHIC REGION: ``Georgia...

78 FR 19066 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel MI CASA; Invitation for...

Science.gov (United States)

2013-03-28

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0035] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel MI CASA; Invitation for Public Comments AGENCY... CASA is: Intended Commercial Use of Vessel: Passenger charters. Geographic Region: Massachusetts, New...

77 FR 35743 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SECOND CHANCE; Invitation for...

Science.gov (United States)

2012-06-14

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2012 0067] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SECOND CHANCE; Invitation for Public Comments AGENCY... SECOND CHANCE is: Intended Commercial Use of Vessel: ``Small group charters.'' Geographic Region...

78 FR 77200 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel AFFAIR; Invitation for Public...

Science.gov (United States)

2013-12-20

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013-0151] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel AFFAIR; Invitation for Public Comments AGENCY: Maritime... AFFAIR is: Intended Commercial Use of Vessel: ``Charter Fishing (sport) & sightseeing tours.'' Geographic...

77 FR 20100 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel NORDIC STAR; Invitation for...

Science.gov (United States)

2012-04-03

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2012 0039] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel NORDIC STAR; Invitation for Public Comments AGENCY... NORDIC STAR is: Intended Commercial Use of Vessel: ``Sailing excursions and extended charters...

78 FR 35091 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BAD INFLUENCE; Invitation for...

Science.gov (United States)

2013-06-11

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0072] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BAD INFLUENCE; Invitation for Public Comments AGENCY....gov . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel BAD...

77 FR 26818 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel FREE SPIRIT; Invitation for...

Science.gov (United States)

2012-05-07

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2012-0058] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel FREE SPIRIT; Invitation for Public Comments AGENCY... SPIRIT is: Intended Commercial Use of Vessel: ``Day/overnight passenger passage. Depart and return same...

78 FR 42153 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel COMPASS ROSE; Invitation for...

Science.gov (United States)

2013-07-15

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013-0081] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel COMPASS ROSE; Invitation for Public Comments AGENCY... COMPASS ROSE is: Intended Commercial Use Of Vessel: ``Sailboat charters six passengers or less...

77 FR 16582 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel KAMI KAY; Invitation for...

Science.gov (United States)

2012-03-21

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2012-0033] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel KAMI KAY; Invitation for Public Comments AGENCY....gov . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel KAMI...

77 FR 58213 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel ISLAND WATERS; Invitation for...

Science.gov (United States)

2012-09-19

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2012 0095] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel ISLAND WATERS; Invitation for Public Comments AGENCY... ISLAND WATERS is: Intended Commercial Use of Vessel: ``Charter to individuals in coastal waters...

78 FR 8698 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel COOL BEANS; Invitation for...

Science.gov (United States)

2013-02-06

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0005] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel COOL BEANS; Invitation for Public Comments AGENCY... BEANS is: Intended Commercial Use of Vessel: Sightseeing and sunset cruises. Geographic Region: Florida...

78 FR 35092 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BEAR BOAT; Invitation for...

Science.gov (United States)

2013-06-11

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013-0069] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BEAR BOAT; Invitation for Public Comments AGENCY....gov . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel BEAR...

76 FR 67253 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel THE GIFT; Invitation for...

Science.gov (United States)

2011-10-31

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2011 0131] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel THE GIFT; Invitation for Public Comments AGENCY... GIFT is: INTENDED COMMERCIAL USE OF VESSEL: ``Passenger carrying.'' GEOGRAPHIC REGION: ``ME, NH, MA, RI...

76 FR 70216 - Requested Administrative Waiver of the Coastwise Trade Laws; Vessel REEL ATTITUDE; Invitation for...

Science.gov (United States)

2011-11-10

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2011 0146] Requested Administrative Waiver of the Coastwise Trade Laws; Vessel REEL ATTITUDE; Invitation for Public Comments AGENCY... ATTITUDE is: Intended Commercial Use of Vessel: ``Fishing Charter passenger less than 6.'' Geographic...

Nuclear reactor vessel surface inspecting technique applying electric resistance probe

International Nuclear Information System (INIS)

Yamaguchi, T.; Enami, K.; Yoshioka, M.

1975-01-01

A new technique for inspecting the inner surface of the PWR type nuclear reactor vessel by use of an electric resistance probe is introduced, centering on a data processing system. This system is composed of a mini-computer, a system typewriter, an interface unit, a D-A converter and controller, and X-Y recorder and others. Its functions are judging flaws and making flaw detection maps. In order to judge flaws by flaw detection signals, three kinds of flaw judging methods have been developed. In case there is a flaw, its position and depth are calculated and listed on the system typewriter. The flaw detection maps are expressed in four kinds of modes and they are displayed on the X-Y recorder. (auth.)

An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

International Nuclear Information System (INIS)

1982-04-01

An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

77 FR 26819 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BLUE PLANET; Invitation for...

Science.gov (United States)

2012-05-07

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2012-0054] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BLUE PLANET; Invitation for Public Comments AGENCY... PLANET is: Intended Commercial use of Vessel: ``Charter to Boy Scouts of America Florida Sea Base...

78 FR 25530 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BLUE DOLPHIN; Invitation for...

Science.gov (United States)

2013-05-01

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013-0049] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BLUE DOLPHIN; Invitation for Public Comments AGENCY... DOLPHIN is: Intended Commercial Use Of Vessel: ``Skippered daysailing in Puget Sound and San Juan Islands...

78 FR 13752 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SPIRIT; Invitation for Public...

Science.gov (United States)

2013-02-28

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0013] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SPIRIT; Invitation for Public Comments AGENCY: Maritime... . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel SPIRIT is: Intended...

77 FR 22631 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SPIRIT; Invitation for Public...

Science.gov (United States)

2012-04-16

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2012 0049] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SPIRIT; Invitation for Public Comments AGENCY: Maritime... . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel SPIRIT is: Intended...

76 FR 76812 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BARBARY GHOST; Invitation for...

Science.gov (United States)

2011-12-08

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2011 0148] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BARBARY GHOST; Invitation for Public Comments AGENCY... BARBARY GHOST is: Intended Commercial Use Of Vessel: ``Sightseeing tours in the San Francisco Bay for a...

78 FR 30961 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel LITTLE DUTCH; Invitation for...

Science.gov (United States)

2013-05-23

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013-0057] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel LITTLE DUTCH; Invitation for Public Comments AGENCY... LITTLE DUTCH is: Intended Commercial Use of Vessel: ``Carrying up to six passengers for day trips, sunset...

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

In-vessel inspection before head removal: TMI II, Phase III (tooling and systems design and verification)

International Nuclear Information System (INIS)

Carter, G.S.; Ryan, R.F.; Pieleck, A.W.; Bibb, H.Q.

1982-09-01

Under EG and G contract K-9003 to General Public Utilities Corporation, a Task Order was assigned to Babcock and Wilcox to develop and provide equipment to facilitate early assessment of core damage in the Three Mile Island Unit 2 reactor vessel head. Described is the work performed, the equipment developed, and the tests conducted with this equipment on various mockups used to simulate the constraints inside and outside the reactor vessel that affect the performance of the inspection. The tooling developed provides several methods of removing a few control rod drive leadscrews from the reactor, thereby providing paths into which cameras and lights may be inserted to permit video viewing of many potentially damaged areas in the reactor vessel. The tools, equipment, and cameras demonstrated that these tasks could be accomplished

78 FR 35093 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel EYE DOC; Invitation for...

Science.gov (United States)

2013-06-11

... Administrative Waiver of the Coastwise Trade Laws: Vessel EYE DOC; Invitation for Public Comments AGENCY... DOC is: INTENDED COMMERCIAL USE OF VESSEL: ``Charter fishing on Lake Erie'' GEOGRAPHIC REGION: ``Ohio..., Maritime Administration. [FR Doc. 2013-13836 Filed 6-10-13; 8:45 am] BILLING CODE 4910-81-P ...

78 FR 70097 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel BEE; Invitation for Public...

Science.gov (United States)

2013-11-22

... Administrative Waiver of the Coastwise Trade Laws: Vessel BEE; Invitation for Public Comments AGENCY: Maritime... entered into this docket is available on the World Wide Web at http://www.regulations.gov . FOR FURTHER....gov . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel BEE...

Inspection device for buried equipment

International Nuclear Information System (INIS)

Hanawa, Jun.

1994-01-01

In an inspection device for a buried equipment, a rail is suspended at the upper portion of a vessel of a pit-vessel type pump buried in a plant building floor, and a truck movable vertical in the vessel along the rail, and an ultrasonic wave probe contained in the truck and urged to the vessel by an electromagnet are disposed. In addition, an elevator moving vertically along a shaft is disposed, and an arm having the ultrasonic probe disposed at the end portion and driven by a piston are disposed to the elevator. The ultrasonic wave probe moves vertically together with the truck along the rail in the vessel while being urged to the vessel by the electromagnet to inspect and measure the state at the inner and outer surfaces of the vessel. Further, the length of the arm is controlled so as to set a predetermined distance between the ultrasonic wave probe and the vessel. Subsequently, the elevator is moved vertically along a shaft passing through a shaft hole of a mount, and the shaft is rotated thereby enabling to inspect and measure the state of the inner and outer surfaces of the vessel. (N.H.)

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)

Operational inspections

International Nuclear Information System (INIS)

Bystersky, M.

1997-01-01

Special equipment is described, designed for inspection of reactor pressure vessels performed from the inside. Central shaft manipulator ZMM-5 is available for crack detection control using ultrasound and eddy currents, for visual check of surfaces, repair works at the reactor pressure vessel, and hardness measurements. The manipulator consists of the manipulator bridge, a cable container, shaft segments, a control mechanism and auxiliary parts. Eight inspections were performed at the Bohunice nuclear power plant and two at the Paks nuclear power plant. (M.D.)

78 FR 19068 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel GOLDEN BOY II; Invitation for...

Science.gov (United States)

2013-03-28

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0036] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel GOLDEN BOY II; Invitation for Public Comments AGENCY... GOLDEN BOY II is: Intended Commercial Use Of Vessel: Limited charter of passengers for luxury day...

78 FR 19362 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel W.L. STEWART III; Invitation...

Science.gov (United States)

2013-03-29

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013 0026] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel W.L. STEWART III; Invitation for Public Comments AGENCY.... STEWART III is: Intended Commercial Use of Vessel: ``Yacht Club Regatta and Sailing Instruction Support...

Development of technology for the in-service inspection of reactor pressure vessel metal condition using a magnetic method

International Nuclear Information System (INIS)

Bakirov, M. B.; Zabruskov, N.Y; Massoud, J.P.

2002-01-01

The opportunity to perform the inspection of condition of base metal and metal of welded joints of cladded vessels of PWR by non destructive methods is shown. The technique for on-site specimen-free testing is offered on the basis of sharing a kinetic hardness method and magnetic method. The results of studies of magnetic and mechanical properties of vessel steels in various condition after irradiation and thermal processing are submitted. It is shown, that the magnetic properties (first of all coercive force) are sensitive to change of structure of a material. (authors)

Emergency disconnect means for the manipulator arm of a nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Jacobs, F.; Morris, D.W.

1980-01-01

An emergency disconnect means is disclosed for uncoupling a portion of the linkage means which connects the several segments of an articulating manipulator arm employed in a nuclear reactor vessel inspection device. One of the motor housings included in the manipulator arm's segmented drive train is pivotably coupled between two segments thereof. In the event of power failure or the necessity of manual retraction of the manipulator arm from within the vessel, a lever is manually operated and moved from its normally locked position wherein the motor housing is positionally fixed to a release position wherein the motor housing and the remainder of the manipulator arm segments connected forwardly thereof are pivotally released to shorten the normal arm reach and alter the normal orientation of the manipulator arm to expedite removal without danger of collision. (auth)

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

77 FR 49059 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel CHAT DE MER; Invitation for...

Science.gov (United States)

2012-08-15

... Administrative Waiver of the Coastwise Trade Laws: Vessel CHAT DE MER; Invitation for Public Comments AGENCY....gov . SUPPLEMENTARY INFORMATION: As described by the applicant the intended service of the vessel CHAT... Francisco Bay. Geographic Region: ``California.'' The complete application is given in DOT docket MARAD-2012...

Nondestructive examination requirements for PWR vessel internals

International Nuclear Information System (INIS)

Spanner, J.

2015-01-01

This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

Development of ultrasonic testing technique with the large transducer to inspect the containment vessel plates of nuclear power plant embedded in concrete

International Nuclear Information System (INIS)

Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

2004-01-01

The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. In order to establish ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely at the accessible point, experiments to detect artificial hollows simulating corrosion on a surface of a carbon steel plate mock-up covered with concrete simulating the embedded containment vessel plates were carried out with newly made ultrasonic transducers. We made newly low frequency (0.3 MHz and 0.5 MHz) surface shear horizontal (SH) wave transducers combined with three large active elements, which were equivalent to a 120mm width element. As a result of the experiments, the surface SH transducers could detect clearly the echo from the hollows with a depth of 9.5 mm and 19 mm at a distance of 1500mm from the transducers on the surface of the mock-up covered with concrete. Therefore, we evaluate that it is possible to detect the defects such as corrosion on the plates embedded in concrete with the newly made low frequency surface SH transducers with large elements. (author)

Inspection and replacement of baffle assembly screws inside American reactor vessels

International Nuclear Information System (INIS)

Neal, K.; Chaumont, J.C.

1999-01-01

The baffle assembly inside the vessel of a 900 MWe reactor designed by Framatome, is made up of 44 plates fixed on 8 horizontal supports by a system of about 1000 screws. These plates undergo high neutron flux and the problem of screw cracking appeared at the end of the eighties in the first-generation reactors. The first operation on a large scale concerning the screws of a Westinghouse type reactor, was performed on the Tihange-1 power plant where Framatome controlled 960 screws and replaced 91. In 1997 as a consequence of the Belgian and French feedback experience, American plant operators launched a vast program of preventive actions: material analysis, inspection of baffle plate screws and replacement of defective screws. This program was held in cooperation with EPRI (electric power research institute) and under the control of NRC (nuclear regulatory commission). Framatome Technologies Inc (FTI) was in charge of the in-situ inspection and replacement of the screws. FTI designed special tools and equipment adapted to the 2-loop American reactors but the basis ideas were those applied on the Tihange reactor. The successful experience of FTI has allowed the firm to be commissioned for 6 2-loops American reactors. (A.C.)

78 FR 23823 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel FISHIN GAME; Invitation for...

Science.gov (United States)

2013-04-22

... will be available for inspection and copying at the above address between 10 a.m. and 5 p.m., E.T... FISHIN GAME is: Intended Commercial Use Of Vessel: Sport fishing excursions with six passengers or less. All fishing will be for sport only. Geographic Region: ``California, Oregon''. The complete...

The PISC exercise: a discussion of its relevance to ultrasonic inspection of pressure vessels

International Nuclear Information System (INIS)

Whittle, M.J.; Coffey, J.M.

1981-01-01

The value of the European Plate Inspection Steering Committee (PISC) exercise for determining the reliability of ultrasonic inspection of pressure vessels is discussed. It is argued that the particular ultrasonic procedure assessed in the exercise was a poor one whose failure was predictable. In addition the test blocks were sufficiently unrealistic that the value of the results would have been diminished even if all the defects had been found. Consequently the report maintains that the outcome of the exercise is largely irrelevant to the wider question of the reliability of more thorough ultrasonic procedures. A more general discussion is given of the role of test block studies in ultrasonics. Statistical arguments are used to show that it is impracticable to produce a sufficient number of defects to demonstrate a high reliability and have a high confidence in such a result. Some important points for planning future programmes are emphasised. (author)

Investigation of in service inspection for pressure vessel of the 200 MW nuclear heating reactor

International Nuclear Information System (INIS)

He Shuyan; Yin Ming; Liu Junjie; Chang Huanjian; Zhou Ningning

1997-01-01

The Nuclear District Heating Reactor (NHR) is a new type of reactor. There are some differences in the arrangement of the primary circuit components and in safety features between NHR and PWR or other reactors. In this paper the safety features of the 200 MW NHR are described. The failure probability, the LBB property and the in-service inspection requirement for the 200 MW NHR pressure vessel are also discussed. (author). 16 refs, 6 figs, 4 tabs

Investigation of in service inspection for pressure vessel of the 200 MW nuclear heating reactor

Energy Technology Data Exchange (ETDEWEB)

Shuyan, He; Ming, Yin; Junjie, Liu; Huanjian, Chang; Ningning, Zhou [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The Nuclear District Heating Reactor (NHR) is a new type of reactor. There are some differences in the arrangement of the primary circuit components and in safety features between NHR and PWR or other reactors. In this paper the safety features of the 200 MW NHR are described. The failure probability, the LBB property and the in-service inspection requirement for the 200 MW NHR pressure vessel are also discussed. (author). 16 refs, 6 figs, 4 tabs.

Development of an automatic reactor inspection system

International Nuclear Information System (INIS)

Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

2002-02-01

Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

Preliminary investigation of ultrasonic shear wave holography with a view to the inspection of pressure vessels

International Nuclear Information System (INIS)

Aldridge, E.E.; Clare, A.B.; Shepherd, D.A.

1975-01-01

The manner in which holography would fit into the general scheme of pressure vessel inspection is discussed. Compared to conventional A, B and C presentations holography requires a different processing of the ultrasonic signal and a mechanical scan which may be more demanding than that normally provided for a C display. Preliminary results are presented of the examination of artificial defects in steel plate using shear wave holography. (author)

Pulley system including emergency locking means for nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Reyes, R.D.

1979-01-01

A pulley system for driving the manipulating arm of a nuclear reactor vessel inspection device is disclosed. Two drive pulleys and pulley take-up spools are disposed in driving engagement with a motor. The pulley cable is looped about a first set of idler pulleys fixed to the manipulator arm which is to be driven along the main or central column of the inspection device. Two additional idler pulleys, a second set, are used to define another portion of the pulley cable loop, which portion serves to equalize unbalanced forces in the cable. When the drive motor is actuated, only that portion of the pulley cable between the drive set and the first set of idler pulleys moves. Depending on the direction of movement, the manipulator arm is raised or lowered with equal force on each side thereof minimizing wear and potential hazards. In addition, to guard against the possible dangers resulting from a snapped pulley cable, a restraining clip is secured to the cable between the second set of idler pulleys and is appropriately sized to jam between one of the idler pulleys and its support bracket. 3 claims

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...

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

PWR vessel flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

1990-01-01

This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

A Laser Metrology/Viewing System for ITER In-Vessel Inspection

International Nuclear Information System (INIS)

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M.; Dagher, M.A.; Slotwinski, A.

1997-10-01

This paper identifies the requirements for a remotely operated precision laser ranging system for the International Thermonuclear Experimental Reactor. The inspection system is used for metrology and viewing, and must be capable of achieving submillimeter accuracy and operation in a reactor vessel that has high gamma radiation, high vacuum, elevated temperature, and magnetic field levels. A coherent, frequency modulated laser radar system is under development to meet these requirements. The metrology/viewing sensor consists of a compact laser-optic module linked through fiberoptics to the laser source and imaging units, located outside the harsh environment. The deployment mechanism is a remotely operated telescopic mast. Gamma irradiation up to 10 7 Gy was conducted on critical sensor components with no significant impact to data transmission, and analysis indicates that critical sensor components can operate in a magnetic field with certain design modifications. Plans for testing key components in a magnetic field are underway

Remote inspection system for components installed inside a primary containment vessel of a boiling water reactor

International Nuclear Information System (INIS)

Shimizu, Katsutoshi; Kawai, Katsumi; Ito, Takahiko; Hashimoto, Yuji; Tomizawa, Fumio.

1983-01-01

A remote operation type monitoring system was developed to always enable the watching of the condition of the main equipment installed in the containment vessels of BWRs. It comprises four inspection vehicles suspended by a monorail and pulled with trolley chain, coaxial cables for signal transmission and power supply, and control system. On the inspection vehicles, a television camera, a thermometer, a microphone and a radiation dose rate meter are installed. The performance of the system was confirmed at 60 deg C for several months. Thereafter, the field test was carried out in the Tokai No. 2 Power Station, Japan Atomic Power Co., from December, 1980, to September, 1981. By the continuous monitoring and grasp of operational condition, the preventive maintenance and the improvement of the rate of operation can be expected. Also it is desirable in view of the reduction of radiation exposure of operators. The mechanization of and the labor saving in inspection and maintenance works is necessary because skilled workers will be short. The design and the composition of the system and its tests are reported. (Kako, I.)

Phased array concept for the ultrasonic inservice inspection of the spherical bottom of BWR-pressure vessels

International Nuclear Information System (INIS)

Brekow, G.; Wuestenberg, H.; Moehrle, W.; Schulz, E.

1989-01-01

The spherical bottom of BWR-pressure vessels contains holes for the nozzles of control rods and instrumentation. Up to now the detectable areas for the ultrasonic inspection are the accessible ligaments between the nozzles with an orientation parallel and transverse to the manipulator rails. Some licensing authorities demand an inspection technique capable of reliably detecting significant crack initiation in all critical areas near the cladding of the spherical inner surface. By order and in cooperation with the HEW we have developed a computer controlled equipment with two ultrasonic probes containing four linear arrays and a digitized A-scan storage for documentation and evaluation of inspection results. The manipulator guided probe movement in the paths between the nozzles of the spherical bottom is controlled by a computer program. This program determines for each array system and for each coupling position the beam angle as a function of the variable skewing angle to realize detection conditions suited to possible crack positions at the longitudinal, transverse and diagonal ligaments between the nozzles for control rods and instrumentation. (orig./HP)

Automatic examination of nuclear reactor vessels with focused search units. Status and typical application to inspections performed in accordance with ASME code

International Nuclear Information System (INIS)

Verger, B.; Saglio, R.

1981-05-01

The use of focused search units in nuclear reactor vessel examinations has significantly increased the capability of flaw indication detection and characterization. These search units especially allow a more accurate sizing of indications and a more efficient follow up of their history. In this aspect, they are a unique tool in the area of safety and reliability of installations. It was this type of search unit which was adopted to perform the examinations required within the scope of inservice inspections of all P.W.R. reactors of the French nuclear program. This paper summarizes the results gathered through the 4l examinations performed over the last five years. A typical application of focused search units in automated inspections performed in accordance with ASME code requirements on P.W.R. nuclear reactor vessels is then described

Competition and safety in the law on technical inspection and control - is it a contradiction?

International Nuclear Information System (INIS)

Roth, H.A.

1987-01-01

For some time now, objections have been raised against the argument that technical inspection and control should indeed remain a task exlusively for the Technical Control Boards and their offices, organisations and employed inspectors, because competition in this very safety-related sector would not be a positive factor of selection but rather a mechanism reducing the inspection quality, resulting in a hazardous lowering of the safety level. The objections are primarily raised by the free-lance technical experts and their organisations who would like to enlarge their field of activity in this direction. The book at hand discusses the question how much free manoeuvering space there is for the legislative body to deal with such demands and reconcile the interests of safety and competition. The problem is discussed with a view to the Basic Law, which is said to create no legal basis for the demand for more competition in this field. The Basic Law leaves this decision to the discretion of the legislature. (orig./HP) [de

Periodic and in-service inspection programs

International Nuclear Information System (INIS)

Dinu, M.

2000-01-01

Periodic and in-service inspection programs for Cernavoda NPP consists of periodic inspections of CANDU NPP components CSAN N-285.4 and CSAN N-285.4, in-service inspections and repair and modifications general inspection. Periodic inspection program document (PIPD) determines the systems and components subject to inspection, the category of the inspection, techniques, areas and other details.The current status of the inspection programs is presented, including containment , erosion/corrosion, pressure vessel support and snubbers, main steam lines inspection programs. Qualification program in Cernavoda NPP involves equipment qualification in the on-site laboratory, yearly certification, special equipment qualification in the National Institute of Metrology. All procedures are approved by the ISCIR (regulatory body for pressure vessel and lifting equipment) and CNCAN (National Commission on Nuclear Activities Control). Qualification of the personnel is performed according to the ISCIR Technical prescription CR 11/82 for up to 3 year period. Final qualification and licensing is performed by CNCAN

Device for remote inspection and testing of a structure

International Nuclear Information System (INIS)

Blanc, B.; Boudou, J.; Castaing, A.; Clasquin, J.; Gallet, B.; Saglio, R.; Samoel, A.

1976-01-01

A self-propelled carriage for inspecting the primary vessel of a fast reactor is capable of displacement within the interspace between the primary vessel and the containment vessel in order to inspect and test any predetermined zone of the primary vessel, the carriage being associated with a drive mechanism and applied against the oppositely facing wall of the containment vessel. The carriage is suspended from a composite cable actuated by a handling apparatus for introducing the carriage into the interspace and withdrawing it therefrom. The composite cable supplies electric power as well as the different fluids required for positioning and operation of the inspection devices which are mounted on the carriage. 9 claims, 6 drawing figures

Advanced inspection technology for non intrusive inspection (NII) program

International Nuclear Information System (INIS)

Zamir Mohamed Daud

2003-01-01

In the current economic environment, plants and facilities are under pressure to introduced cost saving as well as profit maximising measures. Among the many changes in the way things are run is a move towards Risk Based Inspection (RBI), with an emphasis on longer operating periods between shutdowns as well as to utilise components to their maximum capability. Underpinning and RBI program requires good data from an effective online inspections program, which would not require the shutdown of critical components. One methodology of online inspection is known as Non Intrusive Inspection (NII), an inspection philosophy with the objective of replacing internal inspection of a vessel by doing Non Destructive Testing (NDT) and inspections externally. To this end, a variety of advanced NDT techniques are needed to provide accurate online measurements. (Author)

INSPECTION IN THE FIELD OF EMPLOYMENT RELATIONSHIPS AND PENALTY PROVISIONS UNDER THE LABOR LAW

Directory of Open Access Journals (Sweden)

Vojo Belovski

2017-10-01

Full Text Available If the primary efficacy of the projected normative order is absent, the mechanism of secondary efficacy occurs, which implies forceful realization of norms – repressive measures, i.e. coercive measures, and punitive undertakings. In the Labor Law there is a whole Chapter (Chapter XXV devoted to inspection supervision in the field of employment relationships. Labor inspection functions as a specialized organ of the Ministry which is responsible for the affairs in the area of labor. A request for a control by a labor inspectorate can be instigated by a worker individually, by the Trade Union and by the employer. It can also be instigated ex officio by the inspection supervision. Concerning supervision of the legal commencement of employment the procedure for the labor inspector is as follows: a to find employees at the place of work of the employer who are not employed according to the law; b employees not registered in the mandatory social insurance; c shall make a decision and shall order the employer to commence employment with the persons found there or with other persons without public job announcement within 8 days; d to employ for an indefinite period of time; e the number of employees shall not be reduced within the next three months; f will make a proposal for settlement by issuing a payment order of misdemeanor for the person responsible or a person authorized by the employer under the Law on Misdemeanors; g if the employer does not accept the payment order of misdemeanor, the inspector in charge will file a request for initiating a misdemeanor procedure. As an example of the misdemeanor provisions, a fine of 7,000 Euros in denar equivalent shall be pronounced on the employerlegal entity if: 1 if no agreement for employment has been made between the employee and the employer and the employer failed to register the employee in the Mandatory pension and Disability Insurance Fund, Health insurance and insurance in case of unemployment

Analisis Remaining Life dan Penjadwalan Program Inspeksi pada Pressure Vessel dengan Menggunakan Metode Risk Based Inspection (RBI

Directory of Open Access Journals (Sweden)

Dyah Arina Wahyu Lillah

2017-01-01

Full Text Available Seiring perkembangan eksplorasi minyak dan gas bumi di dunia, perusahaan minyak dan gas di Indonesia juga turut berlomba-lomba untuk mendapatkan ladang minyak dan gas bumi sebanyak-banyaknya. Perkembangan ini turut dipengaruhi oleh aturan-aturan pemerintah mengenai keselamatan dan pencegahan bahaya baik pada unit yang dikelola maupun tenaga kerja pengelola. Untuk itu semua perlatan-peralatan (unit kerja harus dijamin kehandalaannya agar tidak menimbulkan bahaya baik bagi pekerja maupun lingkungan. Subjek penelitian dalam tugas akhir ini ialah pada pressure vessel yang dimiliki oleh Terminal LPG Semarang. Kemungkinan bahaya yang dapat menyebabkan kerusakan pada pressure vessel perlu dianalisis agar dapat meminimalkan resiko yang akan terjadi. Metode Risk Based Inspection (RBI diharapkan dapat meminimalkan resiko yang ada pada pressure vessel. Penilaian resiko dalam tugas akhir ini mengacu pada standar API RP 581. Untuk mengetahui besarnya resiko yang ada pada plant, maka terlebih dahulu harus dihitung besarnya probabilitas kegagalan dan konsekuensi apabila terjadi kegagalan. Langkah selanjutnya ialah membandingkan besarnya resiko yang didapat dengan target resiko yang dimiliki oleh perusahaan. Dari hasil perbandingan ini dapat diketahui tingkat resiko pressure vessel, sehingga dapat ditentukan jadwal inspeksi dan metode inspeksi yang tepat.

Design and development of weld inspection manipulator for reactor pressure vessel of TAPS-1

International Nuclear Information System (INIS)

Chatterjee, H.; Singh, J.P.; Ranjon, R.; Kulkarni, M.P.; Patel, R.J.

2013-01-01

The reactor pressure vessel (RPV) of TAPS-1 BWR contains six longitudinal and four circumferential welds. Periodical in-service inspection of these weld joints has been a regulatory issue pending for long. In the 22 nd refuelling outage in July 2012 the inspection of L1-1, L1-2 longitudinal welds as well as their junctions with C1 circumferential weld were proposed to be done using ultrasonic technique. Approaching these welds from OD side of the RPV is a difficult and tedious task. Therefore it was decided to examine these welds from ID side of the RPV by filling the cavity with water and approaching the RPV from top. No technology was locally available to take the probes at a depth of 10-12 m under water. NPCIL approached RTD, BARC to develop an underwater manipulator to accomplish this task. RTD took up this work as a challenge and came out with the design of manipulator. The weld inspection manipulator (WIM) was fabricated on a war foot basis, tested and successfully implemented in the reactor for the first time in TAPS history. The entire activity was completed in three months time. This article gives the details of design, manufacturing, performance testing, qualification trials and implementation of WIM in the reactor. Ultrasonic testing techniques were developed by QAD, BARC which are not covered in this article. (author)

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

Development and operational experiences of an automated remote inspection system for interior of primary containment vessel of a BWR

International Nuclear Information System (INIS)

Ozaki, N.; Chikara, S.; Fumio, T.; Katsuhiro, M.; Katsutoshi, S.; Ken-Ichiro, S.; Masaaki, F.; Masayoshi, S.

1983-01-01

A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests

Inspection control and the environmental protection

Directory of Open Access Journals (Sweden)

Milkov Dragan L.

2015-01-01

Full Text Available Environmental protection is the task of many administrative bodies, but the activity of the environmental inspection is of special importance. According to the Law on Environmental Protection, inspection's tasks in this area belong to the competence of republican bodies, ie. Ministry of Agriculture and Environmental Protection. Autonomous province and local self-government have only delegated competence in this field, under condition that this is explicitly regulated in special laws. Environmental inspection's activity consists in preventive actions, certain prior - preparatory activities and the audit itself. In addition, following the intervention of the Inspector issue of the control may be corrective or repressive. According to the Law on Inspection Control and the Law on Environmental Protection, the inspectors have a number of powers and responsibilities, which enable them to ensure the lawful and proper conduct of individuals and legal entities.

Subaquatic, pressure vessels and LPG storage spheres internal inspection; Inspecao interna de esfera utilizando mergulho como acesso

Energy Technology Data Exchange (ETDEWEB)

Filgueira Filho, Rafael; Monteiro, Ayres [PETROBRAS, Rio de Janeiro, RJ (Brazil)

2005-07-01

Minimizing shut-down costs is a widespread target in the oil and gas industry. The use of new inspection techniques is one of the ways for that. This work presents a new procedure for internal inspections in pressure vessels by the non destructive testing - NDT, ACFM, using industrial diving techniques. As a pioneer experience, this method was applied in the inspection of the internal parts of the LPG sphere tank 5101 at PETROBRAS Transporte S.A. - TRANSPETRO, in Jequie's Terminal, in the state of Bahia, in december, 2003. This new method allows the reduction of indirect costs related to operational unavailability of the equipment, by the reduction of the shut-down time in approximately 50%, when compared to the demanded shut down time, when using scaffolds for accessing the internal parts. Despite of direct costs are still higher with the new methodology, this paper demonstrates the economical feasibility of this new method, based on the savings obtained with the fastest return of the equipment to operation. (author)

Pressurized water reactor inspection procedures

International Nuclear Information System (INIS)

Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

1998-01-01

Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

78 FR 33149 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel CATTITUDE; Invitation for...

Science.gov (United States)

2013-06-03

... Administrative Waiver of the Coastwise Trade Laws: Vessel CATTITUDE; Invitation for Public Comments AGENCY... Floor, Room W12-140, 1200 New Jersey Avenue SE., Washington, DC 20590. You may also send comments electronically via the Internet at http://www.regulations.gov . All comments will become part of this docket and...

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)

Recovery of a broken inspection lance from the reactor vessel of the German sodium cooled fast breeder reactor SNR 300

International Nuclear Information System (INIS)

Menck, J.W.; Hoeft, E.; Kirchner, G.

1988-01-01

An inspection lance for flow and vibration measurements was installed into the SNR-300 rotating plug. Centering and guiding of the lower end of the lance was effected in the central grid plate insert. The lance was torn off due to handling problems. The task consisted in recovering all defective parts from the reactor vessel and re-establishing the intact initial state. (author)

Advanced inspection and repair techniques for primary side components

International Nuclear Information System (INIS)

Elm, Ralph

1998-01-01

The availability of nuclear power plant mainly depends on the components of the Nuclear Steam Supply System (NSSS) such as reactor pressure vessel, core internals and steam generators. The last decade has been characterized by intensive inspection and repair work on PWR steam generators. In the future, it can be expected, that the inspection of the reactor pressure vessel and the inspection and repair of its internals, in both PWR and BWR will be one of the challenges for the nuclear community. Due to this challenge, new, advanced inspection and repair techniques for the vital primary side components have been developed and applied, taking into account such issues as: use of reliable and fast inspection methods, repair of affected components instead of costly replacement, reduction of outage time compared to conventional methods, minimized radiation exposure, acceptable costs. This paper reflects on advanced inspection and repair techniques such as: Baffle Former Bolt inspection and replacement, Barrel Former Bolt inspection and replacement, Mechanized UT and visual inspection of reactor pressure vessels, Steam Generator repair by advanced sleeving technology. The techniques described have been successfully applied in nuclear power plants and improved the operation performance of the components and the NPP. (author). 6 figs

Evaluation of structural welds with respect to reliability and inservice inspection

International Nuclear Information System (INIS)

Marci, G.

1980-01-01

The ongoing developments with respect to large forgings for nuclear pressure vessels and the present philosophy of inservice inspection in the Federal Republic of Germany (FRG) (inspection of weld regions only) seem to be oriented towards a monolithic pressure vessel requiring no inservice inspection. This raises questions as to the task of inservice inspection in the framework of the total reliability of a nuclear pressure vessel. The inherent 'weak spots' associated with welds of ferritic low alloyed steels are the microstructural features of stress relief cracks and regions of coarse grains. An analysis as to the effects of stress relief cracks and regions of coarse grains in the heat affected zones of welds on subcritical crack growth is performed. It is shown that these 'weak spots' cannot be considered a particular safety hazard to vessel integrity. Therefore welds per se should not constitute the criteria for the performance of inservice inspection. Rather, the inservice inspection programme should check the regions where manufacturing history, loading conditions, and materials degradation are expected to enhance subcritical crack growth. (author)

Weld defects analysis of 60 mm thick SS316L mock-ups of TIG and EB welds by ultrasonic inspection for fusion reactor vacuum vessel applications

International Nuclear Information System (INIS)

Buddu, Ramesh Kumar; Shaikh, Shamsuddin; Raole, P.M.; Sarkar, B.

2015-01-01

The present paper reports the weld quality inspections carried with 60 mm thick AISI welds of SS316L. The high thickness steel plates requirement is due to the specific applications in case of advanced fusion reactor structural components like vacuum vessel and others. Different kind welds are proposed for the thick plate joints like Tungsten Inert Gas (TIG) welding, Electron beam welding as per stringent conditions (like very low distortions and residual stresses) for the vacuum vessel fabrication. Mock-ups of laboratory scale welds are fabricated by TIG (multi-pass) and EB (double pass) process techniques and different weld quality inspections are carried by different NDT tests. The welds are examined with Liquid penetrant examination to check sub surface cracks/discontinuities towards the defects observation

In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

International Nuclear Information System (INIS)

Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

2000-01-01

During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

Inspection and repair of reactor pressure vessel (RPV) internals

International Nuclear Information System (INIS)

Bohmann, W.; Poetz, F.; Nicolai, M.

1996-01-01

The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (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)

An overview of in-service inspection for nuclear power plants in Japan

International Nuclear Information System (INIS)

Takano, Eishiro

1996-01-01

In nuclear power stations, for maintaining the safety of machinery and equipment, it is stipulated to carry out periodic inspection. The contents of the periodic inspection and the inspections of important matters are explained. In-service inspection (ISI) is the nondestructive inspection for confirming the soundness of reactor pressure vessels, main pipings, valves and others. The stipulations on the scope and time of the periodic inspection and ISI, the course of introducing ISI into Japan, and the basic way of thinking on ISI are described. Pre-service inspection (PSI) is carried out for the purpose of collecting and recording the initial data for ISI. The range of the objects of nondestructive inspection and leak test and the range of the objects of leak test only are shown. As to nondestructive inspection, the parts to be inspected, the frequency of inspection and the methods of inspection, and the method of leak test are described. As the present state of inspection technology, the automatic ultrasonic flow detectors for reactor vessels, piping welded parts and reactor vessel stud bolts, and the eddy current flaw detector for steam generator tubes are explained. (K.I.)

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

An Articulated Inspection Arm for fusion purposes

Energy Technology Data Exchange (ETDEWEB)

Villedieu, E., E-mail: eric.villedieu@cea.fr [CEA-IRFM, 13108 Saint Paul lez Durance (France); Bruno, V.; Pastor, P.; Gargiulo, L. [CEA-IRFM, 13108 Saint Paul lez Durance (France); Song, Y.T.; Cheng, Y.; Feng, H. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Liu, C. [CEA-IRFM, 13108 Saint Paul lez Durance (France); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Shi, S.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

2016-11-01

Highlights: • Requirements for in vacuum tokamak inspection are presented. • Development of a prototype of the Articulated Inspection Arm is described. • The upgrade of the prototype to convert it into a fully operational device is detailed. • Future applications of inspection robots in the new fusion reactors is discussed. - Abstract: Fusion Tokamaks are complex machines which require special conditions for their operation, in particular, high vacuum inside the vessel and high temperature of the vessel walls. During plasma phases, the first wall components are highly stressed and a control is necessary in case of doubt about their condition. To be able to make safely such an inspection in a short period of time is a great advantage. The Articulated Inspection Arm (AIA) developed by the CEA provides the capability for fast inspections of the first wall overall surface keeping the vacuum and temperature conditions of the vessel. The robot prototype was validated in Tore Supra in 2008. In the frame of a joint laboratory, CEA/IRFM and ASIPP have decided to upgrade the existing AIA prototype to use it routinely in the EAST and WEST tokamaks. The robot has followed an important upgrade program in 2013 and 2014. The document presents the various upgrades made on the mechanics, the sensors, the electronics, the control station and the integration adaptation for the operation on EAST. From the AIA experience, thoughts for future inspection robots are given.

An Articulated Inspection Arm for fusion purposes

International Nuclear Information System (INIS)

Villedieu, E.; Bruno, V.; Pastor, P.; Gargiulo, L.; Song, Y.T.; Cheng, Y.; Feng, H.; Liu, C.; Shi, S.S.

2016-01-01

Highlights: • Requirements for in vacuum tokamak inspection are presented. • Development of a prototype of the Articulated Inspection Arm is described. • The upgrade of the prototype to convert it into a fully operational device is detailed. • Future applications of inspection robots in the new fusion reactors is discussed. - Abstract: Fusion Tokamaks are complex machines which require special conditions for their operation, in particular, high vacuum inside the vessel and high temperature of the vessel walls. During plasma phases, the first wall components are highly stressed and a control is necessary in case of doubt about their condition. To be able to make safely such an inspection in a short period of time is a great advantage. The Articulated Inspection Arm (AIA) developed by the CEA provides the capability for fast inspections of the first wall overall surface keeping the vacuum and temperature conditions of the vessel. The robot prototype was validated in Tore Supra in 2008. In the frame of a joint laboratory, CEA/IRFM and ASIPP have decided to upgrade the existing AIA prototype to use it routinely in the EAST and WEST tokamaks. The robot has followed an important upgrade program in 2013 and 2014. The document presents the various upgrades made on the mechanics, the sensors, the electronics, the control station and the integration adaptation for the operation on EAST. From the AIA experience, thoughts for future inspection robots are given.

77 FR 12843 - Fees for Sanitation Inspections of Cruise Ships

Science.gov (United States)

2012-03-02

... Sanitation Inspections of Cruise Ships AGENCY: Centers for Disease Control and Prevention (CDC), Department... diseases. The fee schedule for sanitation inspections of passenger cruise ships inspected under VSP was... sanitation inspections. These inspections are conducted by CDC's Vessel Sanitation Program (VSP). VSP assists...

Development and application of an LWR reactor pressure vessel-specific flaw distribution

International Nuclear Information System (INIS)

Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.

1991-01-01

Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern 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. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses

47 CFR 80.953 - Inspection and certification.

Science.gov (United States)

2010-10-01

... date on which it is placed in service. (b) An inspection and certification of a ship subject to the... must be recorded in the ship's radiotelephone log and include: (1) The date the inspection was... part and has successfully passed the inspection. (c) The vessel owner, operator, or ship's master must...

78 FR 51728 - Fees for Sanitation Inspections of Cruise Ships

Science.gov (United States)

2013-08-21

... Sanitation Inspections of Cruise Ships AGENCY: Centers for Disease Control and Prevention (CDC), Department... for vessel sanitation inspections for Fiscal Year (FY) 2014. These inspections are conducted by HHS...-yearly inspections and, when necessary, re-inspection. DATES: These fees are effective October 1, 2013...

47 CFR 80.903 - Inspection of radiotelephone installation.

Science.gov (United States)

2010-10-01

... ship passes the inspection, the technician will issue a Communications Act Safety Radiotelephony... 47 Telecommunication 5 2010-10-01 2010-10-01 false Inspection of radiotelephone installation. 80... Passenger Boats § 80.903 Inspection of radiotelephone installation. Every vessel subject to Part III of...

Navigation and vessel inspection circular No. 9-82 change 1. Change 1 to NVIC 9-82 of 10 May 1982, subj: MSD certification. Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-10-08

This Circular revises guidance in Navigation and Vessel Inspection Circular No. 9-82, by providing for acceptance of non-Coast Guard certified sewage treatment plants on foreign flag vessels operating in waters of the United States, if the sewage treatment plants meet the requirements of Annex IV of the International Convention for the Prevention of Pollution from Ships, 1973 (MARPOL). The performance requirements for Annex IV Sewage treatment plants are in Resolution MEPC.2(VI) of the International Maritime Organization (IMO).

46 CFR 176.402 - Initial inspection for certification.

Science.gov (United States)

2010-10-01

..., spars, rigging, sails, piping, main and auxiliary machinery, pressure vessels, steering apparatus... make the vessel available for all applicable inspections discussed in this paragraph, and in Subpart H...

In-pile inspections of the Calder and Chapelcross nuclear reactors

International Nuclear Information System (INIS)

Stewart, G.

1984-01-01

The subject is discussed under the headings: introduction (relevant data about the reactors); inspection policy; photographic inspection (equipment; inspection results (vessel seam welds and plates; top dome welds; top dome internals)); ultrasonic equipment; manipulator; television inspections; concluding remarks. (U.K.)

FINAL DESIGN REVIEW REPORT Subcritical Experiments Gen 2, 3-ft Confinement Vessel Weldment

Energy Technology Data Exchange (ETDEWEB)

Romero, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-28

A Final Design Review (FDR) of the Subcritical Experiments (SCE) Gen 2, 3-ft. Confinement Vessel Weldment was held at Los Alamos National Laboratory (LANL) on September 14, 2017. The review was a focused review on changes only to the confinement vessel weldment (versus a system design review). The changes resulted from lessons-learned in fabricating and inspecting the current set of confinement vessels used for the SCE Program. The baseline 3-ft. confinement vessel weldment design has successfully been used (to date) for three (3) high explosive (HE) over-tests, two (2) fragment tests, and five (5) integral HE experiments. The design team applied lessons learned from fabrication and inspection of these vessel weldments to enhance fit-up, weldability, inspection, and fitness for service evaluations. The review team consisted of five (5) independent subject matter experts with engineering design, analysis, testing, fabrication, and inspection experience. The

Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

Energy Technology Data Exchange (ETDEWEB)

Houry, M., E-mail: Michael.houry@cea.fr [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H. [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Kammerer, N.; Measson, Y. [CEA, LIST, F-92265 Fontenay-aux-Roses (France); Carrel, F.; Schoepff, V. [CEA, LIST, F-91191 Gif-sur-Yvette (France)

2011-10-15

ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

International Nuclear Information System (INIS)

Houry, M.; Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H.; Kammerer, N.; Measson, Y.; Carrel, F.; Schoepff, V.

2011-01-01

ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

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.)

A fracture mechanics and reliability based method to assess non-destructive testings for pressure vessels

International Nuclear Information System (INIS)

Kitagawa, Hideo; Hisada, Toshiaki

1979-01-01

Quantitative evaluation has not been made on the effects of carrying out preservice and in-service nondestructive tests for securing the soundness, safety and maintainability of pressure vessels, spending large expenses and labor. Especially the problems concerning the time and interval of in-service inspections lack the reasonable, quantitative evaluation method. In this paper, the problems of pressure vessels are treated by having developed the analysis method based on reliability technology and probability theory. The growth of surface cracks in pressure vessels was estimated, using the results of previous studies. The effects of nondestructive inspection on the defects in pressure vessels were evaluated, and the influences of many factors, such as plate thickness, stress, the accuracy of inspection and so on, on the effects of inspection, and the method of evaluating the inspections at unequal intervals were investigated. The analysis of reliability taking in-service inspection into consideration, the evaluation of in-service inspection and other affecting factors through the typical examples of analysis, and the review concerning the time of inspection are described. The method of analyzing the reliability of pressure vessels, considering the growth of defects and preservice and in-service nondestructive tests, was able to be systematized so as to be practically usable. (Kako, I.)

BWR vessel and internals project (BWRVIP)

International Nuclear Information System (INIS)

Bilanin, W.J.; Dyle, R.L.

1996-01-01

Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

46 CFR 176.702 - Installation tests and inspections.

Science.gov (United States)

2010-10-01

..., machinery, fuel tank, or pressure vessel is installed aboard a vessel after completion of the initial... 100 GROSS TONS) INSPECTION AND CERTIFICATION Repairs and Alterations § 176.702 Installation tests and...

77 FR 50511 - Fees for Sanitation Inspections of Cruise Ships

Science.gov (United States)

2012-08-21

... Sanitation Inspections of Cruise Ships AGENCY: Centers for Disease Control and Prevention (CDC), Department... schedule for sanitation inspections of passenger cruise ships by VSP was first published in the Federal...), announces fees for vessel sanitation inspections for Fiscal Year (FY) 2013. These inspections are conducted...

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.

MRP-227 Reactor vessel internals inspection planning and initial results at the Oconee nuclear station unit 2

International Nuclear Information System (INIS)

Davidsaver, S.B.; Fyfitch, S.; Whitaker, D.E.; Doss, R.L.

2015-01-01

The U.S. PWR industry has pro-actively developed generic inspection requirements and standards for reactor vessel (RV) internals. The Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) Materials Reliability Program (MRP) has issued MRP-227-A and MRP-228 with mandatory and needed requirements based on the Nuclear Energy Institute (NEI) document NEI 03-08. The inspection and evaluation guidelines contained in MRP-227-A consider eight age-related degradation mechanisms: stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (IASCC), wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling and irradiation growth, and thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep. This paper will discuss the decision planning efforts required for implementing the MRP-227-A and MRP-228 requirements and the results of these initial inspections at the Oconee Nuclear power station (ONS) units. Duke Energy and AREVA overcame a significant technology and NDE challenge by successfully completing the first-of-a-kind MRP-227-A scope requirements at ONS-1 in one outage below the estimated dose and with zero safety issues or events. This performance was repeated at ONS-2 a year later. The remote NDE tooling and processes developed to examine the MRP-227-A scope for ONS-1 and ONS-2 are transferable to other PWRs

46 CFR 176.404 - Subsequent inspections for certification.

Science.gov (United States)

2010-10-01

..., and on a sailing vessel, rigging and sails. The owner or operator must conduct all tests as required by the OCMI, and make the vessel available for all specific inspections and drills required by...

Ultrasonic phased array examination of circumferential weld joint in reactor pressure vessel of BWR

Energy Technology Data Exchange (ETDEWEB)

Nanekar, Paritosh, E-mail: pnanekar@barc.gov.in [Quality Assurance Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jothilakshmi, N. [Quality Assurance Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2013-12-15

Highlights: • Phased array technique developed for weld joint inspection in BWR pressure vessel. • Simulation studies were carried out for conventional and phased array probe. • Conventional ultrasonic test shows in-adequate weld coverage and poor resolution. • Focused sound beam in phased array results in good resolution and sensitivity. • Ultrasonic phased array technique is validated on mock-up with reference defects. - Abstract: The weld joints in the reactor pressure vessel (RPV) of Boiling Water Reactors (BWR) are required to be examined periodically for assurance of structural integrity. Ultrasonic phased array examination technique has been developed in authors’ laboratory for inspection of the top flange to shell circumferential weld joint in RPV of BWRs, which are in operation in India since the late 1960s. The development involved detailed simulation studies for computation of focal laws followed by validation on mock-up. The paper brings out the limitations of the conventional ultrasonic technique and how this can be overcome by the phased array approach for the weld joint under consideration. The phased array technique was successfully employed for field examination of this weld joint in RPV during the re-fuelling outage.

Innovative inspection system for reactor pressure vessels

International Nuclear Information System (INIS)

Mertens, K.; Trautmann, H.

1999-01-01

The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

75 FR 78063 - Passenger Weight and Inspected Vessel Stability Requirements

Science.gov (United States)

2010-12-14

... for determining the maximum passenger capacity of small passenger pontoon vessels either by: Dividing.... 176.110 explicitly calling attention to the OCMI's prerogative to consider a vessel's suitability for... factors to which vessel masters should give special attention in both sections, and a requirement in Sec...

In-service inspection of pressurized water reactors

International Nuclear Information System (INIS)

Rapin, M.; Saglio, R.

1983-01-01

French legislation, which is more demanding than in other countries, had led Electricite de France, the State-owned utility, to acquire better performance in-service inspection facilities than those which existed previously. This fact has spurred the industrial development of the new technical facilities which are used worldwide today. This article presents the ''in-service inspection machine'' (MIS) for the inspection of the welds of a PWR vessel, and the inspection device of steam generator tubes; the MIS allow a remote-viewing, ultrasonic and gamma-graphic inspection; Foucault currents are the only one method adapted to the inspection steam generator tubes [fr

Latest experiences in inspecting the inside of BWR vessel shields

Energy Technology Data Exchange (ETDEWEB)

Alberdi, R.; Gonzalez, E.

2001-07-01

In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

Second Annual Maintenance, Inspection, and Test Report for PAS-1 Cask Certification for Shipping Payload B

International Nuclear Information System (INIS)

KELLY, D.J.

2000-01-01

The Nuclear Packaging, Inc. (NuPac), PAS-1 cask is required to undergo annual maintenance and inspections to retain certification in accordance with U.S. Department of Energy (DOE) Certificate of Compliance USA/9184B(U) (Appendix A). The packaging configuration being tested and maintained is the NuPac PAS-1 cask for Payload B. The intent of the maintenance and inspections is to ensure the packaging remains in unimpaired physical condition. Two casks, serial numbers 2162-026 and 2162-027, were maintained, inspected, and tested at the 306E Development, Fabrication, and Test Laboratory, located at the Hanford Site's 300 Area. Waste Management Federal Services, Inc. (WMFS), a subsidiary of GTS Duratek, was in charge of the maintenance and testing. Cogema Engineering Corporation (Cogema) directed the operations in the test facility. The maintenance, testing, and inspections were conducted successfully with both PAS-1 casks. The work conducted on the overpacks included weighing, gasket replacement, and plastic pipe plug and foam inspections. The work conducted on the secondary containment vessel (SCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, and a helium leak test. The work conducted on the primary containment vessel (PCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, dimensional inspection of the vessel bottom, a helium leak test, and dye penetrant inspection of the welds. The vermiculite material used in the cask rack assembly was replaced

Probabilistic retinal vessel segmentation

Science.gov (United States)

Wu, Chang-Hua; Agam, Gady

2007-03-01

Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

A remote inspection system for use inside reactor containment vessels

International Nuclear Information System (INIS)

Aoki, Toshihiko; Kashiwai, Jun-ichi; Yamamoto, Ikuo; Fukada, Koichi; Yamanaka, Yoshinobu.

1985-01-01

The harsh environment in the reactor-containment vesels of pressurized-water reactor nuclear-power plants precludes the possibility of direct circuit inspection; a remote-inspection system is essential. A robot for performing this task must not only be able to withstand the harsh conditions but must also be small and maneuverable enough to function effectively within complex and confined spaces. The article describes a monorail-type remote-inspection robot developed by Mitsubishi Electric to meet these needs, which is now under trial production and testing. (author)

Annual radioactive waste tank inspection program: 1995

International Nuclear Information System (INIS)

McNatt, F.G. Sr.

1996-01-01

Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1995 to evaluate these vessels and evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report

Annual radioactive waste tank inspection program - 1992

International Nuclear Information System (INIS)

McNatt, F.G.

1992-01-01

Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1992 to evaluate these vessels and evaluations based on data accrued by inspections made since the tanks were constructed are the subject of this report

Annual radioactive waste tank inspection program - 1991

International Nuclear Information System (INIS)

McNatt, F.G.

1992-01-01

Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1991 to evaluate these vessels and evaluations based on data accrued by inspections made since the tanks were constructed are the subject of this report

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...

46 CFR 2.10-120 - Overseas inspection and examination fees.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Overseas inspection and examination fees. 2.10-120... VESSEL INSPECTIONS Fees § 2.10-120 Overseas inspection and examination fees. (a) In addition to any other fee required by this subpart, an overseas inspection and examination fee of $4,585 must be paid for...

An interior vessel viewing system for DIII-D

International Nuclear Information System (INIS)

Senior, R.

1989-11-01

It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

Life extension of the BR2 aluminium vessel

International Nuclear Information System (INIS)

Koonen, E.; Fabry, A.; Chaouadi, R.; Verwerft, M.; Raedt, C. de; Winckel, S. van; Wacquier, W.; Dadoumont, J.; Verwimp, A.

2000-01-01

The BR2 reactor has recently undergone a major refurbishment comprising the replacement of all vessel internals. The vessel itself however was not replaced. An important requalification programme has been executed to prove that the vessel would remain fit during the contemplated life extension period of BR2. Representative material samples could be obtained from the shroud surrounding the vessel. A comprehensive in-service inspection was carried out and a vessel surveillance programme has been established. (author)

Exponential Stabilization of an Underactuated Surface Vessel

Directory of Open Access Journals (Sweden)

Kristin Y. Pettersen

1997-07-01

Full Text Available The paper shows that a large class of underactuated vehicles cannot be asymptotically stabilized by either continuous or discontinuous state feedback. Furthermore, stabilization of an underactuated surface vessel is considered. Controllability properties of the surface vessels is presented, and a continuous periodic time-varying feedback law is proposed. It is shown that this feedback law exponentially stabilizes the surface vessel to the origin, and this is illustrated by simulations.

46 CFR 167.15-10 - Application for annual inspection.

Science.gov (United States)

2010-10-01

... NAUTICAL SCHOOL SHIPS Inspections § 167.15-10 Application for annual inspection. Application in writing for the annual inspection of every nautical school ship required to be inspected by law and the... Inspection, at any local Marine Inspection Office, U.S. Coast Guard, where the nautical school ship may be...

Annual radioactive waste tank inspection program - 1999

International Nuclear Information System (INIS)

Moore, C.J.

2000-01-01

Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1999 to evaluate these vessels and auxiliary appurtenances along with evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report

76 FR 49975 - Inspection of Towing Vessels

Science.gov (United States)

2011-08-11

... Position Indicating Radio Beacon FAST Fatigue Avoidance Scheduling Tool FMCSA Federal Motor Carrier Safety... human sleep and rest requirements. H.R. Conf. Rep. 108-617, 2004 U.S.C.C.A.N. 936, 951. B. Regulatory... owners and managing operators. The Coast Guard has proposed a traditional inspection scheme as one option...

19 CFR 4.71 - Inspection of livestock.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Inspection of livestock. 4.71 Section 4.71 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND SECURITY; DEPARTMENT OF THE TREASURY VESSELS IN FOREIGN AND DOMESTIC TRADES Foreign Clearances § 4.71 Inspection of livestock. A proper export...

International cooperation in production inspections

International Nuclear Information System (INIS)

Limousin, S.

2009-01-01

Nuclear pressure equipment, like the reactor pressure vessel or steam generators, are manufactured in many countries all around the world. As only few reactors were built in the 90's, most of the nuclear safety authorities have lost part of their know how in component manufacturing oversight. For these two reasons, vendor inspection is a key area for international cooperation. On the one hand, ASN has bilateral relationships with several countries (USA, Finland, China...) to fulfill specific purposes. On the other hand, ASN participates in international groups like the MDEP ( Multinational Design Evaluation Program). A MDEP working group dedicated to vendor inspection cooperation enables exchanges of informations (inspection program plan, inspection findings...) among the regulators. Join inspections are organized. International cooperation could lead in the long term to an harmonization of regulatory practices. (author)

Reactor vessel sealing plug

International Nuclear Information System (INIS)

Dooley, R.A.

1986-01-01

This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

International Nuclear Information System (INIS)

Watson, P.S.

1990-01-01

Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

An overview of non destructive inspection services in nuclear power plants

International Nuclear Information System (INIS)

Farley, S.

2004-01-01

Worldwide nuclear power plants are obliged by international and local authorities to perform periodical inspection and maintenance of safety relevant components. Non-Destructive Testing (NDT) techniques such as eddy current, ultrasonic, visual, dye penetrant and radiographic testing have been used and continually developed to inspect a wide range of components and materials. Inspecting such components invariably poses an interesting chal-lenge due to complex component geometries, radiation exposure and the material make-up of the component or its welds. As a leader in services to the nuclear industry, Westinghouse has an immense knowl-edge and experience in inspecting and repairing primary circuit components such as steam generators, reactor vessels, core internals, primary coolant pumps and loops, fuel elements and many other components in hazardous environments. To fulfil the requirements posed by authorities and inspection standards, remotely operated manipulators and vehicles have been designed to bring a diverse variety of probes and cameras to the object of inspection. Each inspection process is tested and qualified by the relevant qualification body. In some cases the results of an inspection may require further in depth analysis or even repair of part of the component. These added challenges have often been met by specifically designed and qualified processes such as for the repair of vessel head penetrations or the re-pair of vessel nozzle safe end welds. This presentation will give a general overview of a range of inspection capabilities and give a few examples in which repair was successfully performed. (author)

Midland reactor pressure vessel flaw distribution

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

1993-12-01

The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center's (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions

Estimated incident cost savings in shipping due to inspections.

Science.gov (United States)

Knapp, Sabine; Bijwaard, Govert; Heij, Christiaan

2011-07-01

The effectiveness of safety inspections of ships has been analysed from various angles, but until now, relatively little attention has been given to translate risk reduction into incident cost savings. This paper provides a monetary quantification of the cost savings that can be attributed to port state control inspections and industry vetting inspections. The dataset consists of more than half a million ship arrivals between 2002 and 2007 and contains inspections of port state authorities in the USA and Australia and of three industry vetting regimes. The effect of inspections in reducing the risk of total loss accidents is estimated by means of duration models, in terms of the gained probability of survival. The monetary benefit of port state control inspections is estimated to range, on average, from about 70 to 190 thousand dollars, with median values ranging from about 20 to 45 thousand dollars. Industry inspections have even higher benefits, especially for tankers. The savings are in general higher for older and larger vessels, and also for vessels with undefined flag and unknown classification society. As inspection costs are relatively low in comparison to potential cost savings, the results underline the importance of determining ships with relatively high risk of total loss. Copyright © 2011 Elsevier Ltd. All rights reserved.

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 ...

Non-destructive inspection using HTS-SQUID on aluminum liner covered by CFRP

International Nuclear Information System (INIS)

Hatsukade, Y.; Yotsugi, K.; Sakaguchi, Y.; Tanaka, S.

2007-01-01

An eddy-current-based SQUID non-destructive inspection (NDI) system to detect deep-lying cracks in multi-layer composite-Al vessels was developed taking advantage of the uncontested sensitivity of HTS-SQUID in low-frequency range. An HTS-SQUID gradiometer was mounted in a pulse tube cryocooler. A pair of differential coils with C-shaped ferrite cores was employed to induce an enhanced eddy current in an Al vessel wrapped in a carbon fiber reinforced plastic (CFRP) cover. Ellipsoidal dome-shaped Al liners containing through cracks, which were made by pressure cycle tests, in the CFRP covers with total thickness of 6 mm (CFPR 3 mm, and Al 3 mm) were inspected by the system. While inducing eddy currents in the vessels with excitation fields at 100 Hz or 7 kHz, the vessels were rotated under the HTS-SQUID. Above the cracks, anomalous signals due to the cracks were clearly detected at both frequencies. These results suggested the SQUID-NDI technique would be a possible candidate for inspection of high-pressure multi-layer composite-Al vessels

Non-destructive inspection using HTS-SQUID on aluminum liner covered by CFRP

Energy Technology Data Exchange (ETDEWEB)

Hatsukade, Y. [Toyohashi University of Technology, 1-1 Hibarigaoka, Tempaku-cho, Toyohashi, Aichi 441-8580 (Japan)], E-mail: hatukade@eco.tut.ac.jp; Yotsugi, K. [Toyohashi University of Technology, 1-1 Hibarigaoka, Tempaku-cho, Toyohashi, Aichi 441-8580 (Japan); Sakaguchi, Y. [SAMTECH Corporation, 1000-18 Enmyo-cho, Kashiwara City, Osaka 582-0027 (Japan); Tanaka, S. [Toyohashi University of Technology, 1-1 Hibarigaoka, Tempaku-cho, Toyohashi, Aichi 441-8580 (Japan)

2007-10-01

An eddy-current-based SQUID non-destructive inspection (NDI) system to detect deep-lying cracks in multi-layer composite-Al vessels was developed taking advantage of the uncontested sensitivity of HTS-SQUID in low-frequency range. An HTS-SQUID gradiometer was mounted in a pulse tube cryocooler. A pair of differential coils with C-shaped ferrite cores was employed to induce an enhanced eddy current in an Al vessel wrapped in a carbon fiber reinforced plastic (CFRP) cover. Ellipsoidal dome-shaped Al liners containing through cracks, which were made by pressure cycle tests, in the CFRP covers with total thickness of 6 mm (CFPR 3 mm, and Al 3 mm) were inspected by the system. While inducing eddy currents in the vessels with excitation fields at 100 Hz or 7 kHz, the vessels were rotated under the HTS-SQUID. Above the cracks, anomalous signals due to the cracks were clearly detected at both frequencies. These results suggested the SQUID-NDI technique would be a possible candidate for inspection of high-pressure multi-layer composite-Al vessels.

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)

Conceptual design studies of in-vessel viewing equipment for ITER

International Nuclear Information System (INIS)

Obara, Kenjiro; Oka, Kiyoshi; Taguchi, Hiroshi; Itoh, Akira; Tada, Eisuke; Shibanuma, Kiyoshi

1996-03-01

In-vessel viewing systems are essential to inspect all surface of in-vessel components so as to detect and locate damages, and to assist in-vessel maintenance operations. The in-vessel viewing operations are categorized into the three cases, which are 1) rapid inspection just after off-normal events such as disruption, 2) scheduled inspection, and 3) supplementary inspection during maintenance operations. In case of the rapid inspection, the viewing systems have to be operated in vacuum (ca. 10 -5 Pa) and high temperature (ca. 300degC) under a gamma ray dose rate of 10 7 R/h. On the other hand, the latter two cases are anticipated to be under atmospheric inert gas, 150degC and 3x10 6 R/h. Accordingly, the in-vessel viewing systems are required to have sufficient durability under those conditions of all cases as well as precision of the vision to all of in-vessel surface. Based on those requirements, scoping studies on various viewing concepts have been performed and the applicability to the ITER conditions have been assessed. As a result, two types of viewing systems have been chosen, which are a periscope type viewing system and a image fiber type viewing system with a multi-joint manipulator. Both systems are based on radiation hard optical elements which are being developed. In this report, the design features of both viewing systems are described, including technical issues for ITER application. Finally, a periscope type viewing system is recommended as a primary system and the following specifications/conditions are proposed for the further engineering design. (1) Unified type periscope with a movable mirror at the tip (2) Integrated lighting device into the periscope (3) Accessed from top vertical ports located at 7.3m from the machine center (4) Proposed configuration with a total length of around 27m and a diameter of 200mm. (author)

Procedures and acceptance criteria for PAS-1 cask inspections

International Nuclear Information System (INIS)

Mercado, J.E.

1998-01-01

The procedures and acceptance criteria that comprise this document were prepared to support a one-time test to certify two PAS-1 casks in accordance with US Department of Energy Certificate of Compliance US A/9184/B(U), which was issued in 1998. The specific inspections addressed in this document are the visual weld inspection and a dimensional inspection of the primary containment vessel

Safeguarding the nuclear safety of WWER-440 reactor pressure vessels at SKODA Plzen

International Nuclear Information System (INIS)

Hrbek, Z.

1986-01-01

The approach is described of the SKODA enterprise to safety assurance and to providing the reliability of WWER-440 reactor pressure vessels. The philosophy is analyzed of in-service inspection and determination of the residual service life of pressure vessels. This follows up on the so-called conception of basic safety whose main aim is to preclude failures at production stage by the selection of suitable material, namely by optimizing the choice of raw materials, of metallurgical procedures such as will lead to high purity of the pressure vessel material, by introducing multiple inspection in production, reducing the sensitivity of materials to technological operations, and by high-quality welds. The quality of in-service inspections is given by the use of technical diagnostic instruments of peak quality and of modern methods of nondestructive materials testing. The instruments and methods used are described. It is stated that the experience gained with in-service inspection will make it possible to draw up operating regulations and safety criteria for nuclear installations and own inspection regulations, this with regard to technical and economic factors. (Z.M.)

Approaching application of risk-based inspection to ASME code section XI

International Nuclear Information System (INIS)

Hedden, Owen F.

1995-01-01

This paper will describe current efforts within the ASME Boiler and Pressure Vessel Committee's Subcommittee on Nuclear Inservice Inspection to introduce risk-based technology to optimize inservice inspection of nuclear power plants. The subcommittee is responsible for the content of ASME Boiler and Pressure Vessel Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. The paper will first provide the historical background for the inspection program currently in Section XI. It will then describe the development of new technology through the ASME Center for Research and Technology Development program. Next, the work now going on in two of the groups under the Section XI committee will be described in detail. Each of these two efforts is directed toward the application of new risk-based inspection technology to nuclear piping systems. Finally, the directions of additional research and applications of the technology will be discussed. (author)

Ultimate load analysis of prestressed concrete reactor pressure vessels considering a general material law

International Nuclear Information System (INIS)

Schimmelpfennig, K.

1975-01-01

A method of analysis is presented, by which progressive fracture processes in axisymmetric prestressed concrete pressure vessels during increasing internal pressure can be evaulated by means of a continuum calculation considering a general material law. Formulations used in the analysis concerning material behaviour are derived on one hand from appropriate results of testing small concrete specimens, and are on the other hand gained by parametric studies in order to solve questions still existing by recalulating fracture tests on concrete bodies with more complex states of stress. Due attention is focussed on investigating the behaviour of construction members subjected to high shear forces (end slabs.). (Auth.)

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)

Inspection and repair of nuclear components

International Nuclear Information System (INIS)

Lahner, K.; Poetz, F.

1993-01-01

Despite careful design, manufacturing and operation, some of the important safety-relevant components show deterioration with time. Because of activation and contamination of these components, their inspection and repair has to be performed with manipulators. Some sophisticated manipulators are described, built by ABB Reaktor and used for inspection, maintenance and repair of PWR steam generators, fuel alignment pins, core baffle former bolts and reactor pressure vessel head penetrations. (Z.S.) 7 figs

Minimum critical crack depths in pressure vessels guidelines for nondestructive testing

International Nuclear Information System (INIS)

Crossley, M.R.; Townley, C.H.A.

1983-09-01

Estimates of the minimum critical depths which can be expected in high quality vessels designed to certain British and American Code rules are given. A simple means of allowing for fatigue crack growth in service is included. The data which are presented can be used to decide what sensitivity and what reporting levels should be employed during an ultrasonic inspection of a pressure vessel. It is emphasised that the minimum crack depths are those which would be relevant to a vessel in which the material is stressed to its maximum permitted value during operation. Stresses may, in practice, be significantly less than this. Less restrictive inspection standards may be established, if it were considered worthwhile to carry out a detailed stress analysis of the particular vessel under examination. (author)

Pulsatile blood flow, shear force, energy dissipation and Murray's Law

Directory of Open Access Journals (Sweden)

Bengtsson Hans-Uno

2006-08-01

Full Text Available Abstract Background Murray's Law states that, when a parent blood vessel branches into daughter vessels, the cube of the radius of the parent vessel is equal to the sum of the cubes of the radii of daughter blood vessels. Murray derived this law by defining a cost function that is the sum of the energy cost of the blood in a vessel and the energy cost of pumping blood through the vessel. The cost is minimized when vessel radii are consistent with Murray's Law. This law has also been derived from the hypothesis that the shear force of moving blood on the inner walls of vessels is constant throughout the vascular system. However, this derivation, like Murray's earlier derivation, is based on the assumption of constant blood flow. Methods To determine the implications of the constant shear force hypothesis and to extend Murray's energy cost minimization to the pulsatile arterial system, a model of pulsatile flow in an elastic tube is analyzed. A new and exact solution for flow velocity, blood flow rate and shear force is derived. Results For medium and small arteries with pulsatile flow, Murray's energy minimization leads to Murray's Law. Furthermore, the hypothesis that the maximum shear force during the cycle of pulsatile flow is constant throughout the arterial system implies that Murray's Law is approximately true. The approximation is good for all but the largest vessels (aorta and its major branches of the arterial system. Conclusion A cellular mechanism that senses shear force at the inner wall of a blood vessel and triggers remodeling that increases the circumference of the wall when a shear force threshold is exceeded would result in the observed scaling of vessel radii described by Murray's Law.

Waste Calcining Facility remote inspection report

International Nuclear Information System (INIS)

Patterson, M.W.; Ison, W.M.

1994-08-01

The purpose of the Waste Calcining Facility (WCF) remote inspections was to evaluate areas in the facility which are difficult to access due to high radiation fields. The areas inspected were the ventilation exhaust duct, waste hold cell, adsorber manifold cell, off-gas cell, calciner cell and calciner vessel. The WCF solidified acidic, high-level mixed waste generated during nuclear fuel reprocessing. Solidification was accomplished through high temperature oxidation and evaporation. Since its shutdown in 1981, the WCFs vessels, piping systems, pumps, off-gas blowers and process cells have remained contaminated. Access to the below-grade areas is limited due to contamination and high radiation fields. Each inspection technique was tested with a mock-up in a radiologically clean area before the equipment was taken to the WCF for the actual inspection. During the inspections, essential information was obtained regarding the cleanliness, structural integrity, in-leakage of ground water, indications of process leaks, indications of corrosion, radiation levels and the general condition of the cells and equipment. In general, the cells contain a great deal of dust and debris, as well as hand tools, piping and miscellaneous equipment. Although the building appears to be structurally sound, the paint is peeling to some degree in all of the cells. Cracking and spalling of the concrete walls is evident in every cell, although the east wall of the off-gas cell is the worst. The results of the completed inspections and lessons learned will be used to plan future activities for stabilization and deactivation of the facility. Remote clean-up of loose piping, hand tools, and miscellaneous debris can start immediately while information from the inspections is factored into the conceptual design for deactivating the facility

The Management Advisory Committee of the Inspection Validation Centre seventh report

International Nuclear Information System (INIS)

1990-07-01

The Management Advisory Committee of the Inspection Validation Centre (IVC/MAC) was set up to review the policy, scope, procedure and operation of the Inspection Validation Centre (IVC), to supervise its operation and to advise and report to the United Kingdom Atomic Energy Authority (UKAEA) appropriately. The IVC was established at the UKAEA Risley Laboratory, to validate the procedures, personnel and equipment proposed by Nuclear Electric for use in the ultrasonic inspection at various stages of the fabrication, erection and operation of the Sizewell 'B' Pressurized Water Reactor (PWR) reactor pressure vessel (RPV) and such other components as are identified by the utility. It is operated by the UKAEA to work as an independent organisation under contract to Nuclear Electric, and results are reported to Nuclear Electric together with the conclusions of the Centre in relation to the validation of individual techniques. At the meetings of the IVC/MAC, the progress on the manufacture of the pressure vessel is also outlined by the PWR Project Director. The vessel has now undergone the final stress relief and post-hydro inspection and is due to be delivered to the Sizewell site before the end of 1990. (author)

Cargo container inspection test program at ARPA's Nonintrusive Inspection Technology Testbed

Science.gov (United States)

Volberding, Roy W.; Khan, Siraj M.

1994-10-01

An x-ray-based cargo inspection system test program is being conducted at the Advanced Research Project Agency (ARPA)-sponsored Nonintrusive Inspection Technology Testbed (NITT) located in the Port of Tacoma, Washington. The test program seeks to determine the performance that can be expected from a dual, high-energy x-ray cargo inspection system when inspecting ISO cargo containers. This paper describes an intensive, three-month, system test involving two independent test groups, one representing the criminal smuggling element and the other representing the law enforcement community. The first group, the `Red Team', prepares ISO containers for inspection at an off-site facility. An algorithm randomly selects and indicates the positions and preparation of cargoes within a container. The prepared container is dispatched to the NITT for inspection by the `Blue Team'. After in-gate processing, it is queued for examination. The Blue Team inspects the container and decides whether or not to pass the container. The shipment undergoes out-gate processing and returns to the Red Team. The results of the inspection are recorded for subsequent analysis. The test process, including its governing protocol, the cargoes, container preparation, the examination and results available at the time of submission are presented.

Advanced cutting, welding and inspection methods for vacuum vessel assembly and maintenance

Energy Technology Data Exchange (ETDEWEB)

Jones, L. E-mail: jonesl@ipp.mgg.de; Alfile, J.-P.; Aubert, Ph.; Punshon, C.; Daenner, W.; Kujanpaeae, V.; Maisonnier, D.; Serre, M.; Schreck, G.; Wykes, M

2000-11-01

ITER requires a 316 l stainless steel, double-skinned vacuum vessel (VV), each shell being 60 mm thick. EFDA (European Fusion Development Agreement) is investigating methods to be used for performing welding and NDT during VV assembly and also cutting and re-welding for remote sector replacement, including the development of an Intersector Welding Robot (IWR) [Jones et al. This conference]. To reduce the welding time, distortions and residual stresses of conventional welding, previous work concentrated on CO{sub 2} laser welding and cutting processes [Jones et al. Proc. Symp. Fusion Technol., Marseilles, 1998]. NdYAG laser now provides the focus for welding of the rearside root and for completing the weld for overhead positions with multipass filling. Electron beam (E-beam) welding with local vacuum offers a single-pass for most of the weld depth except for overhead positions. Plasma cutting has shown the capability to contain the backside dross and preliminary work with NdYAG laser cutting has shown good results. Automated ultrasonic inspection of assembly welds will be improved by the use of a phased array probe system that can focus the beam for accurate flaw location and sizing. This paper describes the recent results of process investigations in this R and D programme, involving five European sites and forming part of the overall VV/blanket research effort [W. Daenner et al. This conference].

Vessel head penetrations: French approach for maintenance in the PLIM program

International Nuclear Information System (INIS)

Champigny, F.

2002-01-01

Full text: In 1991, in the Bugey nuclear power plant, for the first time a leak occurred at the level of a vessel head penetration made with base nickel alloy (Inconel 600). This leak was caused by a primary stress corrosion cracking coming from inside the penetration tube. The crack was trough wall extent and primary fluid went out from the top of the vessel head. Immediately, Electricite de France launched important research programs and expertise in order to understand the root causes and propose solutions to this problem. The root causes confirmed PWSCC, and in the same time solutions for repair were studied and an inspection program was established to check the base metal of other vessel head penetrations. After several tests, repair solutions were abandoned because of their high costs (financial and dosimetry). EDF decided to replace all the vessel heads with Inconel 600 penetrations. Non destructive developments leaded to use eddy currents for detection and characterization but also televisual techniques to confirm. In a second step, in order to inspect without removing the inside thermal sleeve, eddy current and ultrasonic sword probes were achieved and used to inspect all vessel heads penetrations. Up to now, 75% of the vessel head have been replaced on the 900 MW and 1300 MW fleets but to replace wisely the last vessel heads EDF continues to perform NDE of the penetrations on the basis of safety criteria. This paper describes the different steps of the applied policy in France, NDE methods, criteria and the results obtained. (author)

77 FR 18689 - Changes to Standard Numbering System, Vessel Identification System, and Boating Accident Report...

Science.gov (United States)

2012-03-28

... in this final rule have been under discussion with State officials for many years and we think they..., whereby the State could ask the owner of the vessel to visually inspect the visible HIN that is on the... vessel to visually inspect the visible HIN that is on the boat and report the correct information back to...

46 CFR 91.15-1 - Standards in inspection of hulls, boilers, and machinery.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Standards in inspection of hulls, boilers, and machinery... hulls, boilers, and machinery. In the inspection of hulls, boilers, and machinery of vessels, the..., respecting material and inspection of hulls, boilers, and machinery, and the certificate of classification...

46 CFR 189.15-1 - Standards in inspection of hulls, boilers, and machinery.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Standards in inspection of hulls, boilers, and machinery... inspection of hulls, boilers, and machinery. In the inspection of hulls, boilers, and machinery of vessels... chapter, respecting material and construction of hulls, boilers, and machinery, and certificate of...

46 CFR 71.15-1 - Standards in inspection of hulls, boilers, and machinery.

Science.gov (United States)

2010-10-01

... 46 Shipping 3 2010-10-01 2010-10-01 false Standards in inspection of hulls, boilers, and machinery..., boilers, and machinery. In the inspection of hulls, boilers, and machinery of vessels, the standards... and inspection of hulls, boilers, and machinery, and the certificate of classification referring...

Stress analysis of R2 pressure vessel. Structural reliability benchmark exercise

International Nuclear Information System (INIS)

Vestergaard, N.

1987-05-01

The Structural Reliability Benchmark Exercise (SRBE) is sponsored by the EEC as part of the Reactor Safety Programme. The objectives of the SRBE are to evaluate and improve 1) inspection procedures, which use non-destructive methods to locate defects in pressure (reactor) vessels, as well as 2) analytical damage accumulation models, which predict the time to failure of vessels containing defects. In order to focus attention, an experimental presure vessel has been inspected, subjected fatigue loadings and subsequently analysed by several teams using methods of their choice. The present report contains the first part of the analytical damage accumulation analysis. The stress distributions in the welds of the experimental pressure vessel were determined. These stress distributions will be used to determine the driving forces of the damage accumulation models, which will be addressed in a future report. (author)

U.S. and French approaches to reactor pressure vessel integrity

International Nuclear Information System (INIS)

Griesbach, T.J.; Buchalet, C.; Server, W.L.

1990-01-01

The effects of radiation embrittlement on the reactor pressure vessel must be considered for continued safe operation of nuclear power plants. The consequences of radiation embrittlement require detailed assessments of the margins of safety against brittle fracture of the vessel. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and U.S. Regulations often use conservative approaches for these assessments which can eventually lead to severe operational hardships for some plants. Taking a look at alternative integrity approaches, such as those demonstrated in France, could ultimately result in improved ASME Code and Regulatory limits. The French studies have shown the significance of performing proper in- service inspections to reliably show that no defects larger than a predetermined size (or class) exist in the inspected region of a vessel. The predetermined size is based upon previous studies on the types of manufacturing defects which can potentially exist in French vessels. Enhanced linear elastic and elastic-plastic fracture mechanics methodologies can be applied to evaluate such defects to assure that brittle fracture will not occur

9 CFR 381.14 - Inspection concerning purportedly exempted operations.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Inspection concerning purportedly exempted operations. 381.14 Section 381.14 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE... make inspections in accordance with law to ascertain whether any of the provisions of the Act or the...

THE ANALYSIS OF FOREIGN-VESSEL SINKING AS AN EFFORT BY THE GOVERNMENT OF INDONESIA TO COMBAT IUU FISHING PURSUANT TO INTERNATIONAL LAW

Directory of Open Access Journals (Sweden)

Kristiyanto - Kristiyanto

2015-12-01

Full Text Available As an archipelagic state, Indonesia possesses some of the most abundant fishery resources in the world. Geographically, Indonesia’s strategic location makes it a challenge, and it is a shared responsibility for all citizens to preserve and conserve these resources. The strategic location and rich biological as well as non-biological marine resources automatically attract foreign vessels to carry out IUU fishing activities, particularly in the area of ZEEI (Indonesian Exclusive Economic Zone. The Government of Indonesia has taken various preventive measures to combat IUU fishing practices through bilateral cooperations and various laws. In addition, the Government has also taken some repressive efforts by burning and sinking foreign vessels. In this study, the researcher will analyze the governmental action pursuant to international law and examine the extent to which the sinking of the ship is effective from the perspective of international law. This study will be conducted using normative and juridical approach by reviewing and analyzing various national and international legal instruments related to IUU fishing. We hope that this study will be able to deliver theoretical and practical benefits for students and other researchers who are interested in the issue of IUU fishing practices.   Keywords : IUU fishing, marine resources, archipelagic state.

75 FR 74674 - Updates to Vessel Inspection Fees

Science.gov (United States)

2010-12-01

... not more than 300 feet 610 More than 300 feet 955 Freight ships: Length not greater than 100 feet 1,425 More than 100 feet but not more than 300 feet 1,870 More than 300 feet 5,410 Industrial Vessels...): Drill ship MODUs 6,710 [[Page 74676

Babcock experience of automated ultrasonic non-destructive testing of PWR pressure vessels during manufacture

International Nuclear Information System (INIS)

Dikstra, B.J.; Farley, J.M.; Scruton, G.

1990-01-01

Major developments in ultrasonic techniques, equipment and systems for automated inspection have lead, over a period of about ten years, to the regular application of sophisticated computer-controlled systems during the manufacture of nuclear reactor pressure vessels. Ten years ago the use of procedures defined in a code such as ASME XI might have been considered sufficient, but it is now necessary, as was demonstrated by the results of the UKAEA defect detection trials and the PISC II trials, to apply more comprehensive arrays of probes and higher test sensitivities. The ultrasonic techniques selected are demonstrated to be adequate by modelling or test-block exercises, the automated systems applied are subject to stringent quality assurance testing, and very rigorous inspection procedures are used in conjunction with a high degree of automation to ensure reproducibility of inspection quality. The state-of-the-art in automated ultrasonic testing of pressure vessels by Babcock is described. Current developments by the company, including automated flaw recognition, integrated modelling of inspection capability, and the use of electronically scanned variable-angle probes are reviewed. Examples quoted include the automated ultrasonic inspections of the Sizewell B pressurized water reactor vessel. (author)

Development of ultrasonic testing technique with a large transducer to inspect the containment vessel plates embedded in concrete for corrosion on nuclear power plant (2)

International Nuclear Information System (INIS)

Ishida, Hitoshi

2005-01-01

The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. The purpose of this study is establishment of ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely from the accessible point. Experiments to detect artificial hollows simulating corrosion and stud bolts which hold the mold of concrete on a surface of a carbon steel plate mock-up covered with concrete were carried out with newly made low frequency (0.3MHz and 0.5MHz) 90 degrees refraction angle shear horizontal (SH) wave transducers combined with three active elements, which were equivalent to a 120 mm width element. As the results: (1) The echoes from the artificial hollows with a depth of 19 mm and 9.5mm at a distance of 1.5 m and the stud bolts with a diameter of 8mm at a distance of 0.7 - 1.7m could be discriminated clearly. (2) The multiple echoes bouncing three times between the front side and the back side of the plate, which was equivalent to a distance of about 12m, could be discriminated. (3) A divergence angle and a -6dB divergence angle of the large element (combined three elements) transducer were about 7 degrees and about 3 degrees. (4) The echoes from the hollows with a depth of 9.5m could be detected at a distance of 3.6 m with a reflection at the side wall of the mock-up. (5) It was estimated that the maximum distance of detection of the echo from the stud bolt with a diameter of 8mm was about 2.9 ∼ 3.6 m. Therefore we evaluate that the large element transducer can propagate the SH wave to about a half of a distance to the bottom of the embedded containment vessel and it is possible to detect the defects such as corrosion to a distance of 3.6 m. (author)

Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

1990-01-01

The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

ITER vacuum vessel structural analysis completion during manufacturing phase

Energy Technology Data Exchange (ETDEWEB)

Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F. [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Caixas, J.; Fernandez, E.; Zarzalejos, J.M. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Kim, H.-S.; Kim, Y.G. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Privalova, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); and others

2016-11-01

Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

ITER vacuum vessel structural analysis completion during manufacturing phase

International Nuclear Information System (INIS)

Martinez, J.-M.; Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F.; Caixas, J.; Fernandez, E.; Zarzalejos, J.M.; Kim, H.-S.; Kim, Y.G.; Privalova, E.

2016-01-01

Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

Operation of an ITER relevant inspection robot on Tore Supra tokamak

Energy Technology Data Exchange (ETDEWEB)

Gargiulo, Laurent [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)], E-mail: laurent.gargiulo@cea.fr; Bayetti, Pascal; Bruno, Vincent; Hatchressian, Jean-Claude; Hernandez, Caroline; Houry, Michael [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Keller, Delphine [CEA, LIST, Service de Robotique Interactive, F-92265 Fontenay aux Roses (France); Martins, Jean-Pierre [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Measson, Yvan; Perrot, Yann [CEA, LIST, Service de Robotique Interactive, F-92265 Fontenay aux Roses (France); Samaille, Frank [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)

2009-06-15

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years R and D programme was jointly conducted by CEA-LIST Interactive Robotics Unit and the Institute for Magnetic Fusion Research (IRFM) in order to demonstrate the feasibility and reliability of an in-vessel inspection robot relevant to ITER requirements. The Articulated Inspection Arm robot (AIA) is an 8-m long multilink carrier with a payload up to 10 kg operable between plasma under tokamak conditioning environment; its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel. Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration. In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10{sup -6} Pa and 120 deg. C) after a conditioning phase at 200 deg. C to avoid outgassing pollution of the chamber. This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.

A quick guide to API 510 certified pressure vessel inspector syllabus example questions and worked answers

CERN Document Server

Matthews, Clifford

2010-01-01

The API Individual Certification Programs (ICPs) are well established worldwide in the oil, gas, and petroleum industries. This Quick Guide is unique in providing simple, accessible and well-structured guidance for anyone studying the API 510 Certified Pressure Vessel Inspector syllabus by summarizing and helping them through the syllabus and providing multiple example questions and worked answers.Technical standards are referenced from the API 'body of knowledge' for the examination, i.e. API 510 Pressure vessel inspection, alteration, rerating; API 572 Pressure vessel inspection; API

Decree-Law No. 48568 of 4 September 1968 - Competence of the J.E.N. to undertake scientific and technical inspections of private or public bodies involved in nuclear research or nuclear activities

International Nuclear Information System (INIS)

1968-01-01

Under this Decree-Law, the Junta de Energia Nuclear (JEN) is empowered to inspect on a regular basis all establishments and installations involved in nuclear activities. These inspections are made to ensure that the installations are operated safely and efficiently and that radiation protection measures are properly applied. (NEA) [fr

Methodology if inspections to carry out the nuclear outages model

International Nuclear Information System (INIS)

Aycart, J.; Mortenson, S.; Fourquet, J. M.

2005-01-01

Before the nuclear generation industry was deregulated in the United States, refueling and maintenance outages in nuclear power plants usually lasted orotund 100 days. After deregulation took effect, improved capability factors and performances became more important. As a result, it became essential to reduce the critical path time during the outage, which meant that activities that had typically been done in series had to be executed in parallel. The new outage model required the development of new tools and new processes, The 360-degree platform developed by GE Energy has made it possible to execute multiple activities in parallel. Various in-vessel visual inspection (IVVI) equipments can now simultaneously perform inspections on the pressurized reactor vessel (RPV) components. The larger number of inspection equipments in turn results in a larger volume of data, with the risk of increasing the time needed for examining them and postponing the end of the analysis phase, which is critical for the outage. To decrease data analysis times, the IVVI Digitalisation process has been development. With this process, the IVVI data are sent via a high-speed transmission line to a site outside the Plant called Center of Excellence (COE), where a team of Level III experts is in charge of analyzing them. The tools for the different product lines are being developed to interfere with each other as little as possible, thus minimizing the impact of the critical path on plant refueling activities. Methods are also being developed to increase the intervals between inspection. In accordance with the guidelines of the Boiling Water Reactor Vessel and Internals project (BWRVIP), the intervals between inspections are typically longer if ultrasound volumetric inspections are performed than if the scope is limited to IVVI. (Author)

Advanced Demonstration of Motion Correction for Ship-to-Ship Passive Inspections

Energy Technology Data Exchange (ETDEWEB)

Ziock, Klaus-Peter [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Boehnen, Chris Bensing [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ernst, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2013-09-30

Passive radiation detection is a key tool for detecting illicit nuclear materials. In maritime applications it is most effective against small vessels where attenuation is of less concern. Passive imaging provides: discrimination between localized (threat) and distributed (non-threat) sources, removal of background fluctuations due to nearby shorelines and structures, source localization to an individual craft in crowded waters, and background subtracted spectra. Unfortunately, imaging methods cannot be easily applied in ship-to-ship inspections because relative motion of the vessels blurs the results over many pixels, significantly reducing sensitivity. This is particularly true for the smaller water craft where passive inspections are most valuable. In this project we performed tests and improved the performance of an instrument (developed earlier under, “Motion Correction for Ship-to-Ship Passive Inspections”) that uses automated tracking of a target vessel in visible-light images to generate a 3D radiation map of the target vessel from data obtained using a gamma-ray imager.

In-vessel maintenance remote manipulator system

International Nuclear Information System (INIS)

Jimenez, E.

1978-01-01

The radiation environment within the Tokamak Fusion Test Reactor (TFTR) vacuum vessel necessitates the development of a Remote Manipulator System (RMS) to perform required periodic inspection and maintenance tasks. The RMS must be able to perform dexterous operations and handle loads that exceed human capabilities. The limited size of the access ports on the TFTR vacuum vessel and the performance profile, defined by the various handling requirements, present unique design constraints. The design approach and formulation of a RMS configuration which satisfies TFTR requirements is presented herein

Experiments on performance of the multi-layered in-vessel core catcher

International Nuclear Information System (INIS)

Kang, K.H.; Kim, S.B.; Park, R.J.; Cheung, F.B.; Suh, K.Y.; Rempe, J.L.

2004-01-01

LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with a uniform gap of 10 mm. Until now, two types of the in-vessel core catcher were used in this study. The first one is a single layered in-vessel core catcher without an internal coating of the LAVA-GAP-2 test, and the other one is a two layered in-vessel core catcher with a 0.5 mm-thick ZrO 2 internal coating of the LAVA-GAP-3 test. Current LAVA-GAP experimental results indicate that an internally coated in-vessel core catcher has better thermal performance compared with an uncoated in-vessel core catcher. Metallurgical inspections on the test specimens of the LAVA-GAP-3 test have been performed to examine the performance of the coating material and the base carbon steel. Although the base carbon steel had experienced a severe thermal attack to the extent that the microstructures were changed and re-crystallization occurred, the carbon steel showed stable and pure chemical compositions without any oxidation and interaction with the coating layer. In terms of the material aspects, these metallurgical inspection results suggest that the ZrO 2 coating performed well. (authors)

Development of bore tools for pipe inspection

Energy Technology Data Exchange (ETDEWEB)

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

1998-04-01

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

Assessment of the reliability of ultrasonic inspection methods

International Nuclear Information System (INIS)

Haines, N.F.; Langston, D.B.; Green, A.J.; Wilson, R.

1982-01-01

The reliability of NDT techniques has remained an open question for many years. A reliable technique may be defined as one that, when rigorously applied by a number of inspection teams, consistently finds then correctly sizes all defects of concern. In this paper we report an assessment of the reliability of defect detection by manual ultrasonic methods applied to the inspection of thick section pressure vessel weldments. Initially we consider the available data relating to the inherent physical capabilities of ultrasonic techniques to detect cracks in weldment and then, independently, we assess the likely variability in team to team performance when several teams are asked to follow the same specified test procedure. The two aspects of 'capability' and 'variability' are brought together to provide quantitative estimates of the overall reliability of ultrasonic inspection of thick section pressure vessel weldments based on currently existing data. The final section of the paper considers current research programmes on reliability and presents a view on how these will help to further improve NDT reliability. (author)

U.S. Industry Activities on Inspection of CRDM Penetrations

International Nuclear Information System (INIS)

Alley, Tom; Kietzman, Kim; Ammirato, Frank

2002-01-01

The discovery of primary water stress corrosion cracking (PWSCC) in control rod drive mechanisms (CRDM) penetrations in U.S. and European plants prompted the U.S. nuclear industry to focus considerable effort on development and implementation of effective inspection methods. In particular, cracking was discovered in butt welds connecting reactor vessel nozzles to main coolant piping and in control rod drive mechanism (CRDM) head penetration base material and attachment welds. The EPRI Materials Reliability Program (EPRI-MRP) formed an Inspection Committee to address development of industry guidance for inspection of these components, development of effective non-destructive examination (NDE) methods, and demonstration of inspection processes. This paper discusses the MRP activities pertaining to inspection of CRDM penetrations. Results of demonstrations and field inspections conducted will also be summarized. (authors)

The ultrasonic shop map and its use in preservice inspection

International Nuclear Information System (INIS)

Caplan, J.S.

1975-01-01

Prior to the introduction of Section X1 of the ASME Code on Inservice Inspection, a plan was introduced by Westinghouse to perform ultrasonic examinations of areas of high stress and high fluence of reactor pressure vessels in the manufacturer's shop and subsequent to the shop hydrostatic test. The tests provided a shop reference map of ultrasonic responses to use in subsequent preservice and inservice inspections, and attempted to locate any ultrasonic reflections beyond the acceptance standards of ASME Section III and, later, of Section X1. The history of the program is reviewed. Thirty-six vessels were examined during 1970 to 1973. As a result of indications discovered during ultrasonic examination repairs were carried out on five of these. Details are given of inspections and repairs. A summary is also given of the indications detected and of the correlations between the ultrasonic evaluation and actual flow characteristics. (U.K.)

Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Andersen, E.S.; Bowey, R.E.; Diaz, A.A.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1991-01-01

This program is intended to establish the effectiveness, reliability and adequacy of inservice inspection of reactor pressure vessels and primary piping systems and the impact of ISI reliability on system integrity. The objectives of the program include: (a) determine the effectiveness and reliability of ultrasonic inservice inspection (ISI) performed on commercial, light water reactor pressure vessels and piping; (b) recommend Code changes to the inspection procedures to improve the reliability of ISI; (c) using fracture mechanics analysis, determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to assure a suitably low failure probability; (d) evaluate the degree of reliability improvement which could be achieved using improved NDE techniques; and (e) based on importance of component to safety, material properties, service conditions, and NDE uncertainties, formulate improved inservice inspection criteria (including sampling plan, frequency, and reliability of inspection) for revisions to ASME Section XI and regulatory requirements needed to assure suitably low failure probabilities

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

International Nuclear Information System (INIS)

Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

1991-10-01

This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs

46 CFR 176.710 - Inspection and testing prior to hot work.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Inspection and testing prior to hot work. 176.710... testing prior to hot work. (a) An inspection for flammable or combustible gases must be conducted by a... operations involving riveting, welding, burning, or other fire producing actions may be made aboard a vessel...

46 CFR 115.710 - Inspection and testing prior to hot work.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Inspection and testing prior to hot work. 115.710... AND CERTIFICATION Repairs and Alterations § 115.710 Inspection and testing prior to hot work. (a) An... involving riveting, welding, burning, or other fire producing actions may be made aboard a vessel: (1...

Multi-purpose deployer for ITER in-vessel maintenance

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

2015-10-15

Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

Investigation of a weld defect, reactor vessel head Ringhals 2

International Nuclear Information System (INIS)

Embring, G.; Pers-Anderson, E.B.

1994-01-01

During the summer-outage 1993 Ringhals unit 2 vessel head was inspected at weld-area of Alloy 182. One major defect was found Two plus two ''boat-samples'' were taken out from the zone between the weld and the stainless cladding. All samples were sent to Studsviks laboratories for detailed investigations. The metallographic and fractographic investigations revealed that the major weld-defect had been there from manufacturing. The defect was located between the Alloy 182-buttering and the pressure vessel steel SA 533 grB cl 1. No indications of PWSCC or IDSCC were found. An inspection programme was defined. Different types of reference blocks were provided by Ringhals in cooperation with ABB TRC. Reference reflectors of type flat bottom hole (FBH) and eroded notches (EDM), with different sizes and separation were manufactured. One weld sample with manufacturing defects -lack of fusion and slag was inclusions- was present. ABB TRC performed UT inspection in the gap between the penetration and the thermal sleeve. Inspection results like defect identification, defect separation and defect sizing accuracy were compared with result from the destructive inspection. No relevant additional defects were found. An analysing and repair program was performed. A special designed disc sealed off the defect area. (authors). 5 figs., 3 refs

Evaluation of inservice inspection examinatiions

International Nuclear Information System (INIS)

Aldrich, D.A.; Cook, J.F.

1990-05-01

In order to evaluate the effectiveness of Section 11, Division 1, ''Rules for Inservice Inspection of Nuclear Power Plant Components,'' of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code examinations, 26 inservice inspection (ISI) summary reports from 24 facilities were reviewed. It was found that these ASME Code examinations and tests are instrumental in revealing indications and defects in welds and plant components. In addition, this study uncovered that fact that some of the Section 11 requirements are apparently not clear and are misunderstood by some of the facilities. Also, the need for more stringent requirements was evaluated and some Code changes are recommended

In-service inspection of nuclear power-plant pressure components

International Nuclear Information System (INIS)

Lautzenheiser, C.E.

1976-01-01

The early light-water-reactor systems for production of commercial power were designed and fabricated in accordance with the codes then being used for fossil-fired power-generating stations with some design changes for increased inspectability during fabrication. Over the past few years, major strides have been made in in-service inspection technology. Work has been under way to determine the reliability of nondestructive testing methods and to develop formal inspection programs throughout the world. The major problems associated with in-service inspection are the scarcity of qualified personnel, the variability in procedures and data recording between inspection agencies, and exposure of inspection personnel to radiation. Further work will be required to more completely mechanize piping inspections to reduce radiation exposure and to standardize inspection procedures, equipment, and certification of personnel. Worldwide attention to the requirements of the American Society of Mechanical Engineers' Boiler and Pressure Vessel Code, the size and integrity of inspection agencies, and efforts such as the development of personnel qualification and certification guides emphasize the importance of in-service inspection to nuclear safety

Annual radioactive waste tank inspection program -- 1993

International Nuclear Information System (INIS)

McNatt, F.G. Sr.

1994-05-01

Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1993 to evaluate these vessels, and evaluations based on data accrued by inspections made since the tanks were constructed, are the subject of this report. The 1993 inspection program revealed that the condition of the Savannah River Site waste tanks had not changed significantly from that reported in the previous annual report. No new leaksites were observed. No evidence of corrosion or materials degradation was observed in the waste tanks. However, degradation was observed on covers of the concrete encasements for the out-of-service transfer lines to Tanks 1 through 8

The Management Advisory Committee of the Inspection Validation Centre. 2nd report

International Nuclear Information System (INIS)

1985-06-01

The document is the second report of the Management Advisory Committee of the Inspection Validation Centre (I.V.C.). The IVC is concerned with the ultrasonic inspection of the CEGB's proposed PWR reactor pressure vessel, and other components. The report deals with the technical progress since May 1984, and includes: interim validation, retrospective validation, examination of procedures, test assembly manufacture, interim validation of manual forging inspections, and validation facilities. (U.K.)

Evaluation of the integrity of SEP reactor vessels

International Nuclear Information System (INIS)

Hoge, K.G.

1979-12-01

A documented review is presented of the integrity of the 11 reactor pressure vessels covered in the Systematic Evaluation Program. This review deals primarily with the design specifications and quality assurance programs used in the vessel construction and the status of material surveillance programs, pressure-temperature operating limits, and inservice inspection programs of the applicable plants. Several generic items such as PWR overpressurization protection and BWR nozzle and safe-end cracking also are evaluated. The 11 vessels evaluated include Dresden Units 1 and 2, Big Rock Point, Haddam Neck, Yankee Rowe, Oyster Creek, San Onofre 1, LaCrosse, Ginna, Millstone 1, and Palisades

The development of in-cell remote inspection system in Tokai reprocessing plant

International Nuclear Information System (INIS)

Ishibashi, Yuzo

1985-01-01

In the Tokai fuel reprocessing plant, the containment is triple, i.e. the vessel containing radioactive material, then the concrete cell structure and finally the housing building. The fuel reprocessing plant is now proceeding with the development of an in-cell remote inspection system. The inspection system is for inspection of the cell itself and the equipment etc. in the cell, concerning the integrity. Described are the following: the course taken and problems in development of the remote inspection system; development of the floor rambling type remote inspection equipment and the multiple armed type, both for inspection of in-cell ''drip trays''; in-cell equipment inspection devices in specifications etc.; problems in its future development. (Mori, K.)

Review of Section XI inservice inspection program effectiveness

International Nuclear Information System (INIS)

Cook, J.F. Sr.

1993-08-01

To evaluate the effectiveness of Section XI, Division 1, open-quotes Rules for Inservice Inspection of Nuclear Power Plant Components,close quotes of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, searches were performed of the Licensing Event Report and Nuclear Plant Reliability Data System computerized data bases, and a review was made of inservice inspection summary reports. It was found that the Section XI examinations and tests detect flaws in welds and plant components and result in subsequent corrective action. This study also shows that the format and topics of information provided in Section XI-prescribed inservice inspection summary reports vary widely

Review of Section XI inservice inspection program effectiveness

Energy Technology Data Exchange (ETDEWEB)

Cook, J.F. Sr.

1993-08-01

To evaluate the effectiveness of Section XI, Division 1, {open_quotes}Rules for Inservice Inspection of Nuclear Power Plant Components,{close_quotes} of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, searches were performed of the Licensing Event Report and Nuclear Plant Reliability Data System computerized data bases, and a review was made of inservice inspection summary reports. It was found that the Section XI examinations and tests detect flaws in welds and plant components and result in subsequent corrective action. This study also shows that the format and topics of information provided in Section XI-prescribed inservice inspection summary reports vary widely.

Design and manufacture of an ultrasonic inspection device for the friction welds in reactor vessel control rod drive mechanism housings

International Nuclear Information System (INIS)

Cieslav, C.; Peteuil, M.

1985-01-01

The control rod drive mechanism housings of a PWR reactor vessel consist of a stainless steel flange and a Ni-Cr-Fe alloy tube, assembled by friction welding. The properties of the interface and the nature of the adjacent materials require the development of a specific ultrasonic inspection technique which could be easily automated, considering the number of parts involved (77 parts per 1300 MWe reactor vessel). The part has the general shape of a tube (inside diameter: 70 mm, outside diameter: 103 mm). The transition between both forged parent materials (stainless steel/Ni-Cr-Fe alloy) is obtained by a very thin interface, whose general orientation is normal to the tube centerline. The heat affected zone has generally a coarser and more irregular structure than that observed in the parent materials. The design and development were carried out using a prototype machine on test-pieces representative of a control rod drive mechanism housing, and containing the following artificial reflectors: notches obtained by electro-discharge machining on the inside and outside surfaces, on each side of the interface; planar artificial defects, parallel to the interface. These defects, obtained from 2 flat bottomed holes, drilled into the mock-up constituent parts, were conveyed to the interface during friction welding

A simple in-vessel/FW component viewing system for SST-1

International Nuclear Information System (INIS)

Santra, Prosenjit; Biswas, Prabal; Vasava, Kirit R.; Jaiswal, Snehal; Parekh, Tejas; Chauhan, Pradeep; Patel, Hiteshkumar; Pradhan, Subrata

2015-01-01

A simple compact system is being proposed for in-situ visual inspection of around 3800 First Wall (FW) graphite (armour) tiles in the vacuum vessel of SST-1 tokamak. The 2 DOF, manual driven system (permanently stationed inside vacuum vessel behind outer passive stabilizer) at top and bottom mid-plane locations consist of a rack and pinion mechanism operating a arm with a CCD camera/LED mounted on it, moving over a cam profile to cover approximately 1/8 th of the toroidal span of the vacuum vessel both at interior top/bottom locations with in the FW modules. The camera and LED light should withstand the ultrahigh vacuum conditions, prolonged baking temperatures of around 200°C along with high electromagnetic forces inside the vessel. This system can be operated remotely in-between shots from outside the VV through a linear motion feed through providing linear moment to a rack and pinion mechanism connected to the arm. This mechanism provides a better viewing of the inside FW components and vessel wall surface of tokamak with simple engineering and operational effort. Any information can be acquired from system regarding damages to FWC due to interaction with plasma as well as damage of other support structures inside VV. In comparison to more complicated and complex inspection system used in other tokamaks, this mechanism can be used for frequent in vessel visual inspection, which limits the system to be small, simple, occupying less space and custom made. This system is cheap with a minimum time for realization of the concept. The paper will present the conceptual and engineering design aspect of the in-viewing system, CAD images, its advantages and limitations, camera and LED details, data acquisition and the present status of realization of the project. (author)

Validation of Sizewell ''B'' ultrasonic inspections -- Messages for performance demonstration

International Nuclear Information System (INIS)

Conroy, P.J.; Leyland, K.S.; Waites, C.

1994-01-01

At the time that the decisions leading to the construction of the Sizewell ''B'' plant were being made, public concern over the potential hazards of nuclear power was increasing. This concern was heightened by the accident at USA's Three Mile Island plant. The result of this and public pressure was that an extensive public inquiry was held in addition to the UK's normal licensing process. Part of the evidence to the inquiry supporting the safety case relied upon the ability of ultrasonic inspections to demonstrate that the Reactor Pressure Vessel (RPV) and other key components were free from defects that could threaten structural integrity. Evidence from a variety of trials designed to investigate the performance capability of ultrasonic inspection revealed that although ultrasonic inspection had the potential to satisfy this requirement its performance in practice was heavily dependent upon the details of application. It was therefore generally recognized that some form of inspection validation was required to provide assurance that the equipment, procedures and operators to be employed were adequate for purpose. The concept of inspection validation was therefore included in the safety case for the licensing of Sizewell ''B''. The UK validation trials covering the ultrasonic inspections of the Sizewell ''B'' PWR Reactor Pressure Vessel are now nearing completion. This paper summarizes the results of the RPV validations and considers some of the implications for ASME 11 Appendix 8 the US code covering performance demonstration

Apparatus for carrying out ultrasonic inspection of pressure vessels

International Nuclear Information System (INIS)

Dent, K.H.; Challender, R.S.

1975-01-01

A carriage-supported manipulator for taking an ultrasonic scanner mechanism into a coolant nozzle of a nuclear reactor pressure vessel is described. The manupulator is rotatable about the axis of the nozzle and is radially expansible to urge the scanner mechanism into a scanning position within the nozzle

DESIGN OF A WELDING AND INSPECTION SYSTEM FOR WASTE STORAGE CLOSURE

International Nuclear Information System (INIS)

H.B. Smartt; A.D. Watkins; D.P. Pace; R.J. Bitsoi; E.D. Larsen T.R. McJunkin; C.R. Tolle

2005-01-01

This work reported here was done to provide a conceptual design for a robotic welding and inspection system for the Yucca Mountain Repository waste package closure system. The welding and inspection system is intended to make the various closure welds that seal and/or structurally join the lids to the waste package vessels. The welding and inspection system will also perform surface and volumetric inspections of the various closure welds and has the means to repair closure welds, if required. The system is designed to perform these various activities remotely, without the necessity of having personnel in the closure cell

ISO and EIGA standards for cryogenic vessels and accessories

CERN Multimedia

CERN. Geneva

2016-01-01

The EIGA/WG 6’s scope is cryogenic vessels and accessories, including their design, material compatibility, operational requirements and periodical inspection. The specific responsibilities include monitoring international standardization (ISO, CEN) and regulations (UN, TPED, PED...

Laws of physics help explain capillary non-perfusion in diabetic retinopathy.

Science.gov (United States)

Stefánsson, E; Chan, Y K; Bek, T; Hardarson, S H; Wong, D; Wilson, D I

2018-02-01

The purpose is to use laws of physics to elucidate the mechanisms behind capillary non-perfusion in diabetic retinopathy. In diabetic retinopathy, loss of pericytes weakens capillary walls and the vessel dilates. A dilated capillary has reduced resistance to flow, therefore increased flow in that vessel and decreased in adjoining capillaries. A preferential shunt vessel is thus formed from the dilated capillary and the adjacent capillaries become non-perfused. We apply the laws of Laplace and Hagen-Poiseuille to better understand the phenomena that lead to capillary non-perfusion. These laws of physics can give a foundation for physical or mathematical models to further elucidate this field of study. The law of Laplace predicts that a weaker vessel wall will dilate, assuming constant transmural pressure. The Hagen-Poiseuille equation for flow and the Ostwald-de Waele relationship for viscosity predict that a dilated vessel will receive a higher portion of the fluid flow than the adjoining capillaries. Viscosity will decrease in the dilated vessel, furthering the imbalance and resulting in a patch of non-perfused capillaries next to the dilated 'preferential' shunt vessel. Physical principles support or inspire novel hypotheses to explain poorly understood phenomena in ophthalmology. This thesis of pericyte death and capillary remodelling, which was first proposed by Cogan and Kuwabara, already agrees with histological and angiographical observations in diabetic retinopathy. We have shown that it is also supported by classical laws of physics.

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

Ensuring the nuclear safety of VVER-440 reactor pressure vessels in Skoda, Concern Enterprise, Plzen

International Nuclear Information System (INIS)

Hrbek, Z.

1985-01-01

Various types of routine inspections are described of reactor pressure vessels with the aim of identifying residual lifetime and overall safety. The inspection programme includes: choice of systems and instruments, type of tests, test frequency, safety criteria, measures to be taken in case of unsatisfactory results, documentation. The criteria are given for periodical inspections and requirements listed for instruments and equipment. The main three groups of tests are: visual inspection and dimension tests, surface inspection and volumetric inspection. Briefly described is some of the equipment used. (M.D.)

Strategies to reduce PWR inspection time

International Nuclear Information System (INIS)

Guerra, J.; Gonzalez, E.

2001-01-01

During last few years, a constant reduction in inspection time was clearly demanded by most nuclear plant owners. This requirement has to be accomplished without any impact in inspection quality that, in general, has also to be improved. All this in a market with increasing competition that forces price reductions. Under these new demands from our customers, Tecnatom reoriented its development efforts to improve his products and services to meet this challenges. Two of our main inspection activities that have clear impact in outage duration are Steam Generator and Vessel inspections. This paper describes the improvements made in these two activities as an example of the reorientation of our development efforts with a focus on the technical improvements made on the software and robotic tools applied as in the data acquisition and analysis systems. In the Steam Generator inspections, new robots with dual guide tubes are commonly used. New eddy current instruments and software were developed to keep up with the data rates produced by the faster acquisition system. Use of automatic analysis software is also helping to improve speed while reducing cost and improving overall job quality. Production rates are close to double from the previous inspection system. (author)

46 CFR 190.07-90 - Vessels contracted for prior to March 1, 1968.

Science.gov (United States)

2010-10-01

... considered satisfactory so long as they are maintained in good condition to the satisfaction of the Officer in Charge, Marine Inspection. Minor repairs and alterations may be made to the same standards as the... VESSELS CONSTRUCTION AND ARRANGEMENT Structural Fire Protection § 190.07-90 Vessels contracted for prior...

MINAC, A new radiographic inspection capability

International Nuclear Information System (INIS)

Lapides, M.E.; Schonberg, R.

1983-01-01

In service, radiographic inspection of thick section, metallic components has historically been limited to the capabilities of radioisotopic sources such as cobalt and iridium. Although practical in many applications, these sources tend to be restricted by output intensity, by resolution obtainable, and by the physical size and weight of necessary transport shielding. In nuclear plant applications, any radiography is also frequently inhibited by the presence of a radiation background from activated corrosion products on the inspection object which may fog the film before the desired exposure is achieved. In recognition of these limitations and the need for improved capability in the conduct of mandated ten-year inspection of U.S. nuclear plant components such as reactor coolant pumps, pressure vessel nozzles and valve bodies, the Electric Power Research Institute (EPRI) undertook the development of a high energy, X-ray device appropriate for in service inspection of thick section (up to approximately 30 cm of steel) components. The development was conducted by Schonberg Radiation (U.S.A). The product, MINAC, a miniaturized linear accelerator system, has been in use in many plant inspections since April 1981. This paper describes the development and typical applications of MINAC

AECL experience in fuel channel inspection

Energy Technology Data Exchange (ETDEWEB)

Van Drunen, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Gunn, R. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Mayo, W.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, D.A. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

1999-06-01

Inspection of CANDU fuel channels (FC) is performed to ensure safe and economic reactor operation. CANDU reactor FCs have features that make them a unique non-destructive testing (NDT) challenge. The thin, 4 mm pressure-tube wall means flaws down to about 0.1 mm deep must be reliably detected and characterized. This is one to two orders of magnitude smaller than is usually considered of significant concern for steel piping and pressure vessels. A second unique feature is that inspection sensors must operate in the reactor core--often within 20 cm of highly radioactive fuel. Work on inspection of CANDU reactor FCs at AECL dates back over three decades. In that time, AECL staff have provided equipment and conducted or supervised in-service inspections in about 250 FCs, in addition to over 8000 pre-service FCs. These inspections took place at every existing CANDU reactor except those in India and Romania. Early FC inspections focussed on measurement of changes in dimensions (gauging) resulting from exposure to a combination of neutrons, stress and elevated temperature. Expansion of inspection activities to include volumetric inspection (for flaws) started in the mid-1970s with the discovery of delayed hydride cracking in Pickering 3 and 4 rolled joints. Recognition of other types of flaw mechanisms in the 1980s led to further expansion in both pre-service and in-service inspections. These growing requirements, to meet regulatory as well as economic needs, led to the development of a wide spectrum of inspection technology that now includes tests for hydrogen concentration, structural integrity of core components, flaws, and dimensional change. This paper reviews current CANDU reactor FC inspection requirements. The equipment and techniques developed to satisfy these requirements are also described. The paper concludes with a discussion of work in progress in AECL aimed at providing state-of-the-art FC inspection services. (author)

Testing of VVER reactor pressure vessels by TOFD method

International Nuclear Information System (INIS)

Skala, Z.; Vit, J.

2002-01-01

The Time of Flight Diffraction Method (TOFD) - one of the new testing methods capable to obtain the real dimensions of flaws - is presented in the paper.The laboratory experiments on samples with artificial flaws and samples with artificially prepared cracks confirmed the high accuracy of flaw through wall extent sizing by TOFD. This accuracy was confirmed by qualification of methods and systems used by Skoda JS for the in-service inspections of WWER 440 vessel circumferential weld. The qualification also confirmed the ability of TOFD to detect reliably flaws, which can are not reliably detected by standard pulse echo testing. Based on the result of experiments and qualification, the TOFD method shall be used routinely by Skoda JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level

Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

Energy Technology Data Exchange (ETDEWEB)

Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Friconneau, J.P.; Perrot, Y. [CEA Fontenay-aux-Roses, LIST Robotics and Interactive Systems Unit, 92 (France); Palmer, J.D. [Max-Planck-Institut fuer Plasmaphysik Boltzmannstr.2, Garching (Germany)

2003-07-01

The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper.

Articulated inspection arm for ITER, a demonstration in the Tore Supra tokamak

International Nuclear Information System (INIS)

Cordier, J.J.; Gargiulo, L.; Grisolia, C.; Samaille, F.; Palmer, J.D.

2003-01-01

The aim of this program is to demonstrate for ITER the feasibility of an in-vessel remote handling inspection using a long reach, limited payload carrier (1 to 10 kg) for penetration of the ITER chamber through the openings. This device is dedicated to close inspection of the Plasma Facing Components (PFC). An articulated demonstrator called articulated inspection arm (AIA) has been manufactured. A feasibility study of a full AIA operation in Tore Supra was performed, taking into account ITER reference requirements. A scale one demonstration of the AIA under ITER relevant condition is feasible on Tore Supra and would give significant improvement in research results for ITER remote Handling equipment. The test of the AIA demonstrator behaviour is foreseen in 2005 in real Tokamak conditions. The paper presents the full robot concept, the results of the first test campaign, the AIA new design and its integration on Tore Supra. Several potential uses of the AIA for the in vessel components inspection are being studied such as PFC visual inspection, water loop leak testing, laser ablation for wall detritiation and carbon dust and flakes removal are foreseen as utilities to be placed at the AIA head. These various systems are described in the paper

Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

International Nuclear Information System (INIS)

Guidez, J.; Martin, L.; Dupraz, R.

2006-01-01

The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

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

33 CFR 157.150 - Crude Oil Washing Operations and Equipment Manual: Recording information after inspections.

Science.gov (United States)

2010-07-01

... Equipment Manual: Recording information after inspections. 157.150 Section 157.150 Navigation and Navigable... Vessels Inspections § 157.150 Crude Oil Washing Operations and Equipment Manual: Recording information... machines. (2) Revolutions, number of cycles, and length of cycles of each COW machine. (3) Pressure and...

Ultrasonic test results for the reactor pressure vessel of the HTTR. Longitudinal welding line of bottom dome

International Nuclear Information System (INIS)

Nojiri, Naoki; Ohwada, Hiroyuki; Kato, Yasushi

2008-06-01

This paper describes the inspection method, the measured area, etc. of the ultrasonic test of the in-service inspection (ISI) for welding lines of the reactor pressure vessel of the HTTR and the inspection results of the longitudinal welding line of the bottom dome. The pre-service inspection (PSI) results for estimation of occurrence and progression of defects to compare the ISI results is described also. (author)

Design of the reactor vessel inspection robot for the advanced liquid metal reactor

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and Oak Ridge National Laboratory designed a prototype wall-crawling robot to perform weld inspection in an advanced nuclear reactor. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a mock non-hostile environment and shown to perform as expected, as detailed in this report

NRU vessel repair and return to service: enhancing a Canadian R and D asset for the future

Energy Technology Data Exchange (ETDEWEB)

Cox, D.S.; Arnold, J.B.; Lee, J.K., E-mail: coxd@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

The National Research Universal (NRU) reactor was successfully returned to high power operation in 2010 August, after completing extensive inspections and repairs of the calandria vessel, in response to a small leak of heavy water that was discovered in 2009 May. The aluminum alloy vessel material had corroded from the outside surface over many years, and required application of weld build-up and plated weld repair at ten locations, executed by remote tooling deployed from inside the reactor vessel. Many specialized remotely operated tools were developed for inspection, sampling and repair operations. In parallel with the repair efforts, important maintenance activities were performed, enabled by the de-fuelled state with the heavy water drained from the vessel. Two annual inspection cycles have now been completed since the reactor was returned to high power operation, confirming the fitness of the vessel for continued operation. The NRU reactor is now entering its 56th year of operation and the current operating licence extends to 2016. Based on the outcome of a comprehensive Integrated Safety Review, AECL is continuing to implement major equipment upgrades and process improvements to support safe and reliable operation through 2021, for the benefit of Canadians and the world. (author)

Transfer system development for a remote inspection robot in nuclear power plants

International Nuclear Information System (INIS)

Mizuno, M.; Ohnuma, M.; Hamada, K.; Mizutani, T.; Shimada, A.; Segawa, M.; Kubo, K.

1984-01-01

A remote operated robot system has been developed for inspection inside the primary containment vessel (PCV) of nuclear power plants. This system consists of an inspection vehicle, a monorail driving system, a signal transmission system, a power supply system and an operator console.. The system has two main features. First is that the operator can transfer the vehicle at any time from outside the PCV to inside or vice versa through a personnel airlock. The second feature is that the vehicle can be transported from one inspection route to another route at junction points. A prototype inspection robot system was fabricated on a trial basis. Running and inspection performances were confirmed utilizing actual size test apparatus

Nuclear power plant pressure vessels. Control of piping

International Nuclear Information System (INIS)

2000-01-01

The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

Proceedings of a specialist meeting on the ultrasonic inspection of reactor components

International Nuclear Information System (INIS)

1976-01-01

Beside synthesis of two conferences on nondestructive testing and on inspection, the contributions of this conference are reporting experimental observations and research works on ultrasonic techniques, methods, procedures (pre-service or in-service) and equipment for the inspection of nuclear reactor components (pressure vessels, tubing and piping), generally in stainless steel (often austenitic or ferritic) material or in zirconium alloy. Some contributions are also dealing with the relationship between material microstructure and ultrasonic inspection method and equipment, or with the detection and sizing precision of flaws (cracks)

Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

Energy Technology Data Exchange (ETDEWEB)

Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

2001-03-01

The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

Interaction between periodic in-service inspection requirements and design

International Nuclear Information System (INIS)

Prot, A.C.; Saglio, R.

1979-03-01

After reviewing the requirements specific of periodic In-Service Inspection related to safety problems, especially for the pressure vessels, and taking into account the experience gained with several PWR reactors, the authors show these requirements could lead to modify the primary circuit design

Automation and mechanization of in-service inspection of selected equipment in FRG's nuclear power plants

International Nuclear Information System (INIS)

Metke, E.

1988-01-01

The procedures and equipment are described for the automation and mechanization of in-service inspection in nuclear power plants in the FRG, used by the KWU company. Checks of the pressure vessel are done by visual means using a colour tv camera, the method of eddy currents and the ultrasonic method. An analysis is made of the time schedule of ultrasonic inspections, and the central column manipulator is described which allows to check all internal regions of the pressure vessel. Attention is also devoted to other devices, e.g., those for prestressing shanks, cleaning shanks, cleaning thread apertures, etc. A combined probe using the ultrasonic method and the eddy current method serves the inspection of heat exchange tubes in the steam generator. For inspecting the primary circuit the KWU company uses devices for checking and working the inner surface of pipes. Briefly described are examples of using KWU equipment in nuclear power plants in CMEA countries. (Z.M.). 11 figs., 6 refs

Calibration assembly for nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Elsner, H.J.

1981-01-01

A removable calibration assembly can be utilized to verify the angular mounting of transducers in an array employed in an ultrasonic inpsection apparatus, to calibrate one axis of movement of the array with reference to a starting point, or to measure and calibrate the speed per unit of distance of the transducer's ultrasonic beam in the operating medium. The calibration assembly includes both a large and small reflecting surface separated by known distances, and several large cones, the tips of which are machined or adjusted to angles at which certain of the transducers are to be mounted. Clamping means for securing the calibration assembly to the inspection apparatus at a predetermined orientation is provided

40 CFR 59.698 - May EPA enter my facilities for inspections?

Science.gov (United States)

2010-07-01

... PROGRAMS (CONTINUED) NATIONAL VOLATILE ORGANIC COMPOUND EMISSION STANDARDS FOR CONSUMER AND COMMERCIAL..., you may need to help us arrange an inspection with the facility's managers, including clerical support... countries, we expect you to locate them in places where local law does not keep us from inspecting as...

Inspection of complex geometry pieces with an intelligent contact transducer

International Nuclear Information System (INIS)

Chatillon, S.; Roy, O.; Mahaut, St.

2000-01-01

A new multi-element contact transducer has been developed to improve the inspection of components with complex geometry. The emitting surface is flexible in order to optimize the contact with pieces. An algorithm, based on a simplified geometric model, has been used to determine the delays law which allows to control the focal characteristics of the transmitted field. Acquisition data lead in transmission with an articulated transducer validate the behavior provided by simulation. Thus the optimization of the delays law ensures the transmission of a beam which is homogeneous and controlled during the moving of the transducer. Inspections in echo-pulse mode are implemented on a sample simulating a component controlled on site. Results show that the dynamical adaptation of the delays law to the geometry of the piece leads to very good performances

Healthy Lung Vessel Morphology Derived From Thoracic Computed Tomography

Directory of Open Access Journals (Sweden)

Michael Pienn

2018-04-01

Full Text Available Knowledge of the lung vessel morphology in healthy subjects is necessary to improve our understanding about the functional network of the lung and to recognize pathologic deviations beyond the normal inter-subject variation. Established values of normal lung morphology have been derived from necropsy material of only very few subjects. In order to determine morphologic readouts from a large number of healthy subjects, computed tomography pulmonary angiography (CTPA datasets, negative for pulmonary embolism, and other thoracic pathologies, were analyzed using a fully-automatic, in-house developed artery/vein separation algorithm. The number, volume, and tortuosity of the vessels in a diameter range between 2 and 10 mm were determined. Visual inspection of all datasets was used to exclude subjects with poor image quality or inadequate artery/vein separation from the analysis. Validation of the algorithm was performed manually by a radiologist on randomly selected subjects. In 123 subjects (men/women: 55/68, aged 59 ± 17 years, the median overlap between visual inspection and fully-automatic segmentation was 94.6% (69.2–99.9%. The median number of vessel segments in the ranges of 8–10, 6–8, 4–6, and 2–4 mm diameter was 9, 34, 134, and 797, respectively. Number of vessel segments divided by the subject's lung volume was 206 vessels/L with arteries and veins contributing almost equally. In women this vessel density was about 15% higher than in men. Median arterial and venous volumes were 1.52 and 1.54% of the lung volume, respectively. Tortuosity was best described with the sum-of-angles metric and was 142.1 rad/m (138.3–144.5 rad/m. In conclusion, our fully-automatic artery/vein separation algorithm provided reliable measures of pulmonary arteries and veins with respect to age and gender. There was a large variation between subjects in all readouts. No relevant dependence on age, gender, or vessel type was observed. These data may

VISA-2, Reactor Vessel Failure Probability Under Thermal Shock

International Nuclear Information System (INIS)

Simonen, F.; Johnson, K.

1992-01-01

1 - Description of program or function: VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjust on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. 2 - Method of solution: The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation. 3 - Restrictions on the complexity of the problem: Maxima of 30 welds. VISA2 models only the belt-line (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only

Developments for the ultrasonic in-service inspection of the inner near surface zone of circular welds in the vessel of the fast breeder-reactor SNR 300

International Nuclear Information System (INIS)

Salzburger, H.J.; Huebschen, G.; Neuschwander, R.; Kleffner, G.; Wessels, J.

1990-01-01

For in-service, ultrasonic inspection of the inner near-surface zone of the welds 1 and 6 of the vessel of the SNR-300, an automated system has been developed for scanning, manipulation and evaluation tasks. The scanning system which currently is tested and qualified as a prototype system uses heat-resistant EMUS probes which do not require coupling agents, and which have been optimized for three different testing tasks. The required electronic equipment has also been developed to prototype stage, and both have been incorporated into the automated manipulation, data acquisition and evaluation system. (orig.) [de

46 CFR 54.01-2 - Adoption of division 1 of section VIII of the ASME Boiler and Pressure Vessel Code.

Science.gov (United States)

2010-10-01

... Boiler and Pressure Vessel Code. 54.01-2 Section 54.01-2 Shipping COAST GUARD, DEPARTMENT OF HOMELAND... division 1 of section VIII of the ASME Boiler and Pressure Vessel Code. (a) Pressure vessels shall be designed, constructed, and inspected in accordance with section VIII of the ASME Boiler and Pressure Vessel...

Track type ultrasonic inspection device

International Nuclear Information System (INIS)

Kajiyama, Shigeru; Sasaki, Tsukasa; Takahisa, Kazuo.

1993-01-01

The present invention concerns an improvement of a scanning device disposed near an object to be inspected such as a nuclear pressure vessel and having an ultrasonic probe, mounted thereon that travel along a running track. Specifically, one of wheel supports on both sides is attached being secured to the scanning device. The other of the supports is capable of fixing and releasing, as well as providing and releasing pressure to and from wheels upon mounting and detachment. This enables to provide a structure capable of pressing the wheels of the running device to the plane of the track and release thereof. Accordingly, it is possible to improve the running performance, reduce the size and weight and shorten the time for mounting and detachment of the running inspection device. (I.S.)

Development of automatic reactor vessel inspection systems; development of data acquisition and analysis system for the nuclear vessel weld

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Po; Park, C. H.; Kim, H. T.; Noh, H. C.; Lee, J. M.; Kim, C. K.; Um, B. G. [Research Institute of KAITEC, Seoul (Korea)

2002-03-01

The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine heavy vessel welds. In order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet. In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed. In this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition software was developed. The new systems were tested on the RPV welds of Ulchin Unit 6 to confirm their functions and capabilities. They worked very well as designed and the tests were successfully completed. 13 refs., 34 figs., 11 tabs. (Author)

Inspection of nuclear reactor welding by acoustic emission

International Nuclear Information System (INIS)

Prine, D.W.

1977-01-01

The objective of the work described is to evaluate in-process acoustic emission weld monitoring with the goal of upgrading the inspection techniques for assuring better weld quality in nuclear reactor piping and pressure vessels. To accomplish this overall objective, the following specific goals have been set within a three year program: (1) prove the feasibility of in-process AE under shop conditions and validate the AE findings by currently acceptable NDE techniques; (2) develop and build a prototype monitor for use in Nuclear Fabrication shops; and (3) provide data for NRC/ASME acceptance of the inspection practice

Development and commissioning of a links type inspection manipulator

International Nuclear Information System (INIS)

Whitcomb, H.M.; Ford, G.T.

1988-01-01

In order to access a number of tortuous inspection routes in the Heysham II/Torness reactors, the Central Electricity Generating Board (CEGB) Remote Inspection Group (RIG) have developed a 'links' type manipulator in conjunction with Babcock Power Ltd. This manipulator operates from the reactor pile cap and allows placement of a television camera within the pressure vessel up to 4m laterally from a standpipe of only 140mm in diameter. The design features of this novel manipulator are described together with the solutions developed to overcome particular difficulties experienced during commissioning. (author)

Crack growth rates in vessel head penetration materials

International Nuclear Information System (INIS)

Gomez Briceno, D.; Lapena, J.; Blazquez, F.

1994-01-01

The cracks detected in reactor vessel head penetrations in certain European plants have been attributed to Primary Water Stress Corrosion Cracking (PWSCC). The penetrations in question are made from Inconel 600. The susceptibility of this alloy to PWSCC has been widely studied in relation to use of this material for steam generator tubes. When the first reactor vessel head penetration cracks were detected, most of the available data on crack propagation rates were from test specimens made from steam generator tubes and tested under conditions that questioned the validity of these data for assessment of the evolution of cracks in penetrations. For this reason, the scope of the Spanish Research Project on the Inspection and Repair of PWR reactor vessel head penetrations included the acquisition of data on crack propagation rates in Inconel 600, representative of the materials used for vessel head penetrations. (authors). 1 fig., 2 tabs., 6 refs

Manipulator for pressure vessel open at the top

International Nuclear Information System (INIS)

Bauer, R.; Kastl, H.

1985-01-01

A manipulator is provided, which has a mast, which can be fixed inside the reactor pressure vessel with a support surrounding the mast which can be moved along the mast for a carrier, which can turn around the mast and is provided with a measuring, testing, inspection or repair device. (orig./HP) [de

Development and operation of the Above Dome Inspection Rig (ADIR)

International Nuclear Information System (INIS)

Dickson, R.P.; Moorby, J.

1984-01-01

Hinkley Point 'B' is developing its remote inspection equipment in order to be able to inspect reactor internals adequately without manned vessel entry. The Above Dome Inspection Rig has been built to allow a number of inspection systems to be introduced and operated within the reactor. The ability to introduce, use and remove inspection equipment without the necessity to lift the rig from the reactor is a vital feature in the speed achieved in completing inspections quickly. Television was selected for the A.D.M. because it has significant advantages in terms of operational convenience. However the quality of image obtained in terms of information available compares unfavourably with photography. The sharpness of a photographic image is largely dictated by the chemical structure of the emulsion, whereas video is limited by the picture line structure and bandwidth. The need for a photographic system for in reactor use is therefore essential for high definition inspection requirements. The first inspection system that has been developed for the ADIR is the Telefilm camera. It consists of a Hasselblad photographic camera using an Insight television camera looking through its viewfinder. The characteristics of television and photography have been combined. (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

Inspection, evaluation and maintenance guidelines for reactor vessel internals in JAPAN

International Nuclear Information System (INIS)

Sakashita, Akihiro; Goto, Tomoya; Hirano, Shinro; Dozaki, Koji

2010-01-01

Inspection and Evaluation Guidelines for reactor internals has been taken into the Rules on Fitness-for–Service for Nuclear Power Plants of The Japan Society of Mechanical Engineers. It is a base of the maintenance plan of each Nuclear Power Plant. A plant maintenance methodology will have more importance to maintain the plant safety and stable plant operation. This paper introduces the systematization of the maintenance such as repair, replacement, preventive maintenance in these guidelines. Maintenance methodologies are classified follows. Repair: methodology to reinforce degraded parts by some methods or prevent progress of degradation of without replacement of the existing structure when the degradation of structure is actualized. Replacement: methodology to replace the existing structure with new one when the degradation of structure is actualized. Preventive maintenance : methodology to mitigate the damaged condition. When the maintenance methodologies are implemented in the actual plant, we have to consider the feedback of the inspection program and plant life management. (author)

Short and long term maintenance strategy for reactor vessel head penetrations

International Nuclear Information System (INIS)

Teissier, A.; Heuze, A.

1995-01-01

This paper presents elements based on : surveys, operating inspection, theoretical studies, safety analysis, laboratory results, that enabled to determine maintenance options and short and long term strategies for processing on reactor vessel head leaks. (TEC). 1 tab

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.)

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

Development of an ITER relevant inspection robot

Energy Technology Data Exchange (ETDEWEB)

Gargiulo, L.; Bayetti, P.; Cordier, J.J.; Grisolia, C.; Hatchressian, J.C. [Association Euratom-CEA, Cadarache (France). Dept. de Recherche sur la Fusion Controlee; Friconneau, J.P.; Keller, D.; Perrot, Y. [CEA-LIST Robotics and Interactive Systems Unit, Fontenay aux Roses (France)

2007-07-01

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in vessel inspection operations without loss of conditioning could be very useful. Within this framework, the aim of the project called AIA (Articulated Inspection Arm) is to demonstrate the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 11 degrees of freedom and a total range of 8 m. The project is currently developed by the CEA within the European workprogramme. Its first in situ tests are planned this summer on the Tore Supra tokamak at Cadarache (France). They will validate chosen concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for generic application. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is currently being manufactured and will allow for close visual inspection of the complex Plasma Facing Components (limiters, neutralisers, RF antennae, diagnostic windows, etc.). - In situ localisation of leakage based on helium sniffer is also studied to improve maintenance operations. - Finally the laser ablation system for PFC detritiation, also developed in CEA laboratories, is being fitted to be implanted into the robot and put into operation in Tore Supra. This paper deals with the integration of the robot in the Tore Supra tokamak and the advances in the development of the listed processes. It also introduces the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions. (orig.)

Development of an ITER relevant inspection robot

Energy Technology Data Exchange (ETDEWEB)

Gargiulo, Laurent [Association Euratom-CEA, Departement de Recherche sur la Fusion Controlee, CE Cadarache 13108 (France)], E-mail: laurent.gargiulo@cea.fr; Bayetti, Pascal; Bruno, Vincent; Cordier, Jean-Jacques [Association Euratom-CEA, Departement de Recherche sur la Fusion Controlee, CE Cadarache 13108 (France); Friconneau, Jean-Pierre [CEA-LIST Robotics and Interactive Systems Unit, CE Fontenay Aux Roses (France); Grisolia, Christian; Hatchressian, Jean-Claude; Houry, Michael [Association Euratom-CEA, Departement de Recherche sur la Fusion Controlee, CE Cadarache 13108 (France); Keller, Delphine; Perrot, Yann [CEA-LIST Robotics and Interactive Systems Unit, CE Fontenay Aux Roses (France)

2008-12-15

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in-vessel inspection operations without loss of conditioning will be mandatory. In this context, an Articulated Inspection Arm (AIA) is currently developed by the CEA within the European work programme framework, which aims at demonstrating the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 8 degrees of freedom and a total range of 8 m. The first in situ tests will take place by the end of 2007 on the Tore Supra Tokamak at Cadarache (France). They will validate concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for various applications. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is already manufactured and will allow close visual inspection of the complex Plasma Facing Components (PFC) (limiters, neutralisers, RF antenna, diagnostic windows, etc.). - In situ localisation of water leakage based on a helium sniffing system is also being studied to improve and facilitate maintenance operations. - Finally a laser ablation system for PFC detritiation, developed in CEA laboratories, is being fitted to be implemented on the robot for future operation in Tore Supra. This paper deals with the integration of the robot into Tore Supra and the progress in the development of the processes listed above. It also describes the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions.

Development of an ITER relevant inspection robot

International Nuclear Information System (INIS)

Gargiulo, L.; Bayetti, P.; Cordier, J.J.; Grisolia, C.; Hatchressian, J.C.

2007-01-01

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in vessel inspection operations without loss of conditioning could be very useful. Within this framework, the aim of the project called AIA (Articulated Inspection Arm) is to demonstrate the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 11 degrees of freedom and a total range of 8 m. The project is currently developed by the CEA within the European workprogramme. Its first in situ tests are planned this summer on the Tore Supra tokamak at Cadarache (France). They will validate chosen concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for generic application. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is currently being manufactured and will allow for close visual inspection of the complex Plasma Facing Components (limiters, neutralisers, RF antennae, diagnostic windows, etc.). - In situ localisation of leakage based on helium sniffer is also studied to improve maintenance operations. - Finally the laser ablation system for PFC detritiation, also developed in CEA laboratories, is being fitted to be implanted into the robot and put into operation in Tore Supra. This paper deals with the integration of the robot in the Tore Supra tokamak and the advances in the development of the listed processes. It also introduces the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions. (orig.)

Development of an ITER relevant inspection robot

International Nuclear Information System (INIS)

Gargiulo, Laurent; Bayetti, Pascal; Bruno, Vincent; Cordier, Jean-Jacques; Friconneau, Jean-Pierre; Grisolia, Christian; Hatchressian, Jean-Claude; Houry, Michael; Keller, Delphine; Perrot, Yann

2008-01-01

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In particular, in-vessel inspection operations without loss of conditioning will be mandatory. In this context, an Articulated Inspection Arm (AIA) is currently developed by the CEA within the European work programme framework, which aims at demonstrating the feasibility of a multi-purpose in-vessel Remote Handling inspection system using a long reach, limited payload carrier (up to 10 kg). It is composed of 5 segments with 8 degrees of freedom and a total range of 8 m. The first in situ tests will take place by the end of 2007 on the Tore Supra Tokamak at Cadarache (France). They will validate concepts for operations under ITER relevant vacuum and temperature conditions. After qualification, the arm will constitute a promising tool for various applications. Several processes are already considered for ITER maintenance and will be demonstrated on the AIA robot carrier: - The first embedded process is the viewing system. It is already manufactured and will allow close visual inspection of the complex Plasma Facing Components (PFC) (limiters, neutralisers, RF antenna, diagnostic windows, etc.). - In situ localisation of water leakage based on a helium sniffing system is also being studied to improve and facilitate maintenance operations. - Finally a laser ablation system for PFC detritiation, developed in CEA laboratories, is being fitted to be implemented on the robot for future operation in Tore Supra. This paper deals with the integration of the robot into Tore Supra and the progress in the development of the processes listed above. It also describes the current test campaign aiming to qualify the robot performance and reliability under vacuum and temperature conditions

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

Development of an inspection robot under iter relevant vacuum and temperature conditions

Energy Technology Data Exchange (ETDEWEB)

Hatchressian, J-C; Bruno, V; Gargiulo, L; Bayetti, P; Cordier, J-J; Samaille, F [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA Cadarache, F-13108 Saint Paul-Lez-Durance Cedex (France); Keller, D; Perrot, Y; Friconneau, J-P [CEA, LIST, Service de Robotique Interactive, 18 route du Panorama, BP6, Fontenay aux Roses F-92265 France (France); Palmer, J D [EFDA-CSU Max-Planck-Institut fuer Plasma Physik Boltzmannstr.2, D-85748 Garching Germany (Germany)

2008-03-15

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. In vessel inspection operations without loss of conditioning could be very mandatory. Within this framework, the aim of the Articulated Inspection Arm (AIA) project is to demonstrate the feasibility of a multi-purpose in-vessel Remote Handling inspection system. It is a long reach, composed of 5 segments with in all 8 degrees of freedom, limited payload carrier (up to 10kg) and a total range of 8m. The project is currently developed by the CEA within the European work program. Some tests will validate chosen concepts for operations under ITER relevant vacuum and temperature conditions. The presence of magnetic fields, radiation and neutron beams will not be considered. This paper deals with the choices of the materials to minimize the out-gassing under vacuum and high temperature during conditioning, the implantation of the electronics which are enclosed in boxes with special gaskets, the design of the first embedded process which is a viewing system.

Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

International Nuclear Information System (INIS)

Heddleson, F.A.

1975-01-01

Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

An ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Kumar, Anish, E-mail: anish@igcar.gov.in; Rajkumar, K.V.; Sharma, Govind K.; Dhayalan, R.; Jayakumar, T.

2015-02-15

Highlights: • We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology. - Abstract: The paper presents a novel ultrasonic methodology developed for in-service inspection (ISI) of shell weld of core support structure of main vessel of 500 MWe prototype fast breeder reactor (PFBR). The methodology comprises of the inspection of shell weld immersed in sodium from the outsider surface of the main vessel using a normal beam longitudinal wave ultrasonic transducer. Because of the presence of curvature in the knuckle region of the main vessel, the normal beam longitudinal wave enters the support shell plate at an angle and forms the guided waves by mode conversion and multiple reflections from the boundaries of the shell plate. Hence, this methodology can be used to detect defects in the shell weld of the core support structure. The successful demonstration of the methodology on a mock-up sector made of stainless steel indicated that an artificial defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably.

Investigation with automatic ultrasonic equipment to trace flaws in a large test piece, and experience gained in carrying out inspections

International Nuclear Information System (INIS)

Lindner, J.P.

1975-01-01

Based on the FRG codes providing guide lines for the Reactor Safety Commission regarding the size and location of flaws to be detected during in-service inspections, investigations were carried out into the possibility of detecting defects in thick-walled reactor pressure vessel components with the aid of ultrasonic inspection systems. A large test rig was used and, in a similar manner to the in-service inspections on a reactor, the tests were carried out with remote-controlled, automatically guided inspection equipment. For this purpose, a test specimen weighing about 10 tons was produced and provided with two weld seams having a large number of artificial defects. Essential parameters for the various reflectors in the test specimen were the size, location, angle and roughness or structure of the reflecting surfaces. As it is known that austenitic cladding has a considerable influence on flaw detection, the tests were undertaken first without cladding and then with cladding. A manipulator was designed for automatic remote-controlled inspection with which the inspection system travels on a meandering route over the area to be inspected. The inspection system employed was of the same type as the one used for baseline tests during external inspections of reactor vessel walls with parallel surfaces. Digital data collection was by a magnetic tape recorder designed to store both the data of the ultrasonic inspection system as well as the allied position data. The data stored on the tape are evaluated with electronic data processing programmes especially developed for this purpose. These programmes allow locally coherent indication patterns to be prepared, thus simplifying the interpretation of the data obtained. The author initially describes the equipment with the aid of which the studies were undertaken. A detailed discussion is then presented on the design of the test specimen and the inspection systems employed. Following this, the results obtained are explained and

Remote surface testing and inspection vehicle

International Nuclear Information System (INIS)

Hyde, E.A.; Goldsmith, H.A.; Proudlove, M.J.

1981-01-01

A remotely controlled vehicle capable of roving over the outer surface of a nuclear reactor primary vessel carrying inspection instrumentation. The vehicle comprises an elongate bridge having a pair of suction support pads. Each pad carries gas thrusters for acting in opposition to the suction effort thereby to reduce adherence of the pads and enable displacement of the vehicle over the surface. The vehicle is supported by a services conducting umbilical. (author)

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

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

Ultrasound and Eddy-Current inspection of longitudinal shroud welds in the Santa Maria de Garona Nuclear Power Plant

International Nuclear Information System (INIS)

Elorza, J.; Fernandez, F.

1998-01-01

This is the first time that this kind of inspection has been performed in the world with the Ultrasound (UT) and Eddy-Current (EC) techniques. The inspection, performed during the 19th refueling outage in March 1997, using the MIDAS acquisition and analysis system, which allows joint UT and EC inspection. MIDAS is the new inspection system developed by Tecnatom for vessel internals that integrates acquisition and evaluation of the data obtained from the use of Ultrasound and Eddy-Current inspection techniques. The inspection was completed in less than the scheduled time, and covered the programmed volumes without any noteworthy incidents. (Author)

Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

International Nuclear Information System (INIS)

Gargiulo, L.; Cordier, J.J.; Friconneau, J.P.; Grisolia, C.; Palmer, J.D.; Perrot, Y.; Samaille, F.

2007-01-01

The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools

Towards operations on Tore Supra of an ITER relevant inspection robot and associated processes

Energy Technology Data Exchange (ETDEWEB)

Gargiulo, L. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)], E-mail: laurent.gargiulo@cea.fr; Cordier, J.J. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Friconneau, J.P. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Grisolia, C. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Palmer, J.D. [EFDA CSU, Max-Planck-Institut fuer Plasma Physik Boltzmannstr. 2, D-85748 Garching (Germany); Perrot, Y. [CEA-LIST Robotics and Interactive Systems Unit, BP6 F-92265 Fontenay aux Roses Cedex (France); Samaille, F. [Association Euratom-CEA, DSM/Departement de Recherche sur la Fusion Controlee, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)

2007-10-15

The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools.

33 CFR 157.148 - COW system: Evidence for inspections.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false COW system: Evidence for... CARRYING OIL IN BULK Crude Oil Washing (COW) System on Tank Vessels Inspections § 157.148 COW system... inspector evidence that the COW system has been installed in accordance with the plans accepted under § 157...

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

Reliability aspects of radiation damage in reactor pressure vessel mterials

International Nuclear Information System (INIS)

Brumovsky, M.

1985-01-01

The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

Remote controlled stud bolt handling device for reactor pressure vessel

International Nuclear Information System (INIS)

Shindo, Takenori; Shigehiro, Katsuya; Ito, Morio; Okada, Kenji

1988-01-01

In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

More recent developments for the ultrasonic testing of light water reactor pressure vessels

International Nuclear Information System (INIS)

Seiger, H.; Engl, G.

1976-01-01

The development of an ultrasonic testing method for the inspection from the outside of the areas close to the cladding of the spherical fields of holes of light water reactor pressure vessels is described

Lessons learned from the NRU vessel leak repair and return to service projects

International Nuclear Information System (INIS)

Heeney, P.; Turcotte, J.

2011-01-01

In May 2009 the National Research Universal (NRU) reactor was shut down due to a small leak detected from the reactor vessel into the annulus surrounding the reactor. What ensued was a challenging, yet successful, 15 month long Repair and Return to Service Outage. This Repair and Return to Service Outage presented many first-of-a-kind challenges that provide learning opportunities which have been incorporated into subsequent planned outages. These lessons learned are invaluable tools to be used in the planning and execution of future outages. Following the repair of the NRU vessel, AECL was required to conduct annual inspections of the vessel wall. These inspections require an annual Extended Outage (up to 4 weeks in length). A planned Extended Outage was conducted in May/June 2011 and provided an opportunity to implement some of the lessons learned during the Repair and Return to Service Outage. Lessons learned from that Extended Outage have been incorporated in the subsequent monthly maintenance outages, with lessons learned sessions being held after each outage to ensure that the execution of outages is constantly improving. (author)

Experience of in-service surveillance and monitoring of prestressed concrete pressure vessels for nuclear reactors

International Nuclear Information System (INIS)

Irving, J.; Smith, J.R.; Eadie, D.McD.; Hornby, I.W.

1976-01-01

Details are given of the statutory requirements for the inspection of prestressed concrete pressure vessels in the United Kingdom, with particular emphasis on the prestressing system. The results of periodic examinations under the Licencing Conditions of the Oldbury and Wylfa vessels are presented and discussed in relation to design expectations and future requirements. Strain, moisture and temperature records obtained from the Oldbury PCPV's over a 10 year period are compared with prediction and new developments in vessel instrumentation are discussed. (author)

Pressure vessels dossier restoration according to NR-13 requirements; Enquadramento de vasos de pressao a norma NR-13

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Jose L. [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil); Goncalves, Osorio C. [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil)

2005-07-01

Pressure vessels are static pressurized equipment typical in oil industry facilities. In TRANSPETRO terminals and stations as well as in the whole PETROBRAS, these equipment can be found in the form of condenser accumulators, separators, heat exchangers, storage spheres and others. Because they work sustaining pressure and, many times flammable fluids, pressure vessels have a reasonable potential for hazard. For this reason, the NR-13 regulation was created. It deals with the safety in maintenance, operation and inspection of pressure vessels and boilers. During the compliance to the NR- 13 rules, a problem usually found is the lack of documents for different reasons. In this case, the NR-13 obligates the owner to recreate the vessel documentation under the responsibility of a chartered professional. This paper presents a case study where NR-13 rules were conformed by tasks involving documentation reconstruction based on information collected by means of inspection and tests performed on the field. (author)

The Management Advisory Committee of the Inspection Validation Centre - fifth report

International Nuclear Information System (INIS)

1988-07-01

The Management Advisory Committee of the Inspection Validation Centre (IVC/MAC) was set up by the Chairman of the UKAEA early in 1983 with terms of reference to review the policy, scope, procedure and operation of the Inspection Validation Centre, to supervise its operation and to advise and report to the UKAEA appropriately. The Inspection Validation Centre (IVC) has been established at the UKAEA Northern Research Laboratories, Risley for the purpose of validating the procedures, equipment and personnel proposed by the CEGB for use in the ultrasonic inspection at various stages of the fabrication, erection and operation of the CEGB's PWR reactor pressure vessel and such other components as are identified by the CEGB. This report, for 1987/8, states that the IVC has continued to make progress in the provision of the validation services as specified. (author)

French administrative practice and design codes for nuclear vessels

International Nuclear Information System (INIS)

Roche, R.L.

1987-07-01

French regulations on boilers and pressure vessels have prevailed for a very long time, the first measure having been promulgated on 29 October 1823. Restraining the attention to nuclear pressure vessels it must be pointed out regulations and enforcement by public authorities are more stringent than they are for conventional pressure vessels. The first part of this paper will be devoted to regulations with a special attention to the decree of 26 February 1974 and to the practice of public authorities in this field with special attention given to the Bureau de Controle de la Construction Nucleaire (BCCN = Bureau of Inspection of Nuclear Design and Manufacturing). The second part of this paper will deal with the French construction codes for nuclear components RCC-M (water reactors) and RCC-MR (elevated temperature design)

3. report of the Management Advisory Committee of the Inspection Validation Centre

International Nuclear Information System (INIS)

1986-07-01

The Inspection Validation Centre (IVC) has been established at the UKAEA Risley Nuclear Power Development Laboratories for the purpose of validating procedures, equipment and personnel proposed by the CEGB for use in the ultrasonic inspection at different stages of the fabrication, erection and operation of the CEGB's proposed PWR pressure vessel and such other components as are identified by the CEGB. Technical progress since May 1985 is reported. The number of operators receiving certificates for detection and sizing of flaws is given. (author)

Role of radiation embrittlement in reactor vessel integrity assessment

International Nuclear Information System (INIS)

Marston, T.U.; Chexal, V.K.; Wyckoff, M.

1982-01-01

Reactor vessel integrity calculations are complex. The effect of radiation embrittlement on vessel material properties is a very important aspect of any vessel integrity evaluation. The importance of realistic (based on surveillance capsule results) rather than conservative estimates of the material properties (based on regulatory curves) cannot be overestimated. It is also important to make realistic thermal hydraulic and system operations assumptions. In addition, use of actual flaw sizes from in-service inspections (versus hypothetical flaw size selection) will promote realism. Important research results exist that need to be incorporated into the regulatory process. The authors believe results from current research and development efforts will demonstrate that, with reasonable assumptions and best estimate calculations, the safety of even the older reactor vessels with high copper content welds can be assured over their design lifetimes without the need for major fixes. The utilities, through EPRI and the vendors, have dedicated a significant effort to solving the pressurized thermal shock problem

New developments for optimization of automatic inspections in nuclear power plants

International Nuclear Information System (INIS)

Bollini, G.J.; Gonzalez, B.

1989-01-01

One of the most important factors to be taken into acount with respect to remote control mechanized inspection tasks carried out during refueling outages in nuclear power plants, which include components and systems as varied as the reactor vessel, nozzles, piping, vessel head closure bolts, turbine discs, etc., is the unification of systems used to control the inspection systems themselves. Likewise, the implementation in these systems of specific functions simplifying operational sequences and confirming correct execution of work implies a set of criteria that must be taken into account if the reliability of the tasks carried out is to be increased. The system presented in this paper, SIROCCO (Sistema Robotizado de Control) has been designed taking into consideration the above mentioned criteria, with two clear obectives in mind: a reduction in inspection times and in the dose received by the participating technical personnel, and an increase in the reliability of the electronic equipment used. In February and March of this year, during shutdown of a BMR-type nuclear power plant, several SIROCCO control units were usids in mechanized inspection systems of various kinds, applicable to components such as those mentioned above, with forecasts regarding the reduction of inspection times being met. The main reasons for this reduction were the familiarity of the operators with this single control system, the unification of replacement parts for equipment maintenance and the availability of interchangeable units. This philosophy of unification of electronic systems also includes data acquisition and the transmission of information between different areas, which require the unification of control systems already commented and accomplished. This course of action is currently under development at Tecnatom. (author). 4 figs

Report of the DHS Small Vessel Security Institute

Science.gov (United States)

2007-10-19

Local Government Interests: LtCol Don Holway; Florida Fish and Wildlife Conservation Commission Maj John Fetterman , Maine Department of Marine...AWW program to help fill gaps in law enforcement patrols. 68 National Small Vessel Security Summit Major John C. Fetterman – State of Maine...Major Fetterman currently serves as Vice President of the National Association of State Boating Law Administrators and for the last 30 years has been a

Risk Based Inspection of Gas-Cooling Heat Exchanger

Directory of Open Access Journals (Sweden)

Dwi Priyanta

2017-09-01

Full Text Available On October 2013, Pertamina Hulu Energi Offshore North West Java (PHE – ONWJ platform personnel found 93 leaking tubes locations in the finfan coolers/ gas-cooling heat exchanger. After analysis had been performed, the crack in the tube strongly indicate that stress corrosion cracking was occurred by chloride. Chloride stress corrosion cracking (CLSCC is the cracking occurred by the combined influence of tensile stress and a corrosive environment. CLSCC is the one of the most common reasons why austenitic stainless steel pipework or tube and vessels deteriorate in the chemical processing, petrochemical industries and maritime industries. In this thesis purpose to determine the appropriate inspection planning for two main items (tubes and header box in the gas-cooling heat exchanger using risk based inspection (RBI method. The result, inspection of the tubes must be performed on July 6, 2024 and for the header box inspection must be performed on July 6, 2025. In the end, RBI method can be applicated to gas-cooling heat exchanger. Because, risk on the tubes can be reduced from 4.537 m2/year to 0.453 m2/year. And inspection planning for header box can be reduced from 4.528 m2/year to 0.563 m2/year.

Civacuve analysis software for mis machine examination of pressurized water reactor vessels

International Nuclear Information System (INIS)

Dubois, Ph.; Gagnor, A.

2001-01-01

The product software CIVACUVE is used by INTERCONTROLE for the analysis of UT examinations, for detection, performed by the In-Service Inspection Machine (MIS) of the vessels of nuclear power plants. This software is based on an adaptation of an algorithm of SEGMENTATION (CEA CEREM), which is applied prior to any analysis. It is equipped with tools adapted to industrial use. It allows to: - perform image analysis thanks to advanced graphic tools (Zooms, True Bscan, 'contour' selection...), - backup of all data in a database (complete and transparent backup of all informations used and obtained during the different analysis operations), - connect PC to the Database (export of Reports and even of segmented points), - issue Examination Reports, Operating Condition Sheets, Sizing curves... - and last, perform a graphic and numerical comparison between different inspections of the same vessel. Used in Belgium and France on different kind of reactor vessels, CIVACUVE has allowed to show that the principle of SEGMENTATION can be adapted to detection exams. The use of CIVACUVE generates a important time gain as well as the betterment of quality in analysis. Wide data opening toward PC's allows a real flexibility with regard to client's requirements and preoccupations

Quality assuring measures for pressure vessels - system approaches, certification, accreditation, surveillance

International Nuclear Information System (INIS)

Link, M.

1992-01-01

Quality assurance measures for pressure vessels in accordance with German codes and standards and with the participation of manufacturers, plant operators and third party inspection agencies represent a high standard in terms of engineering, safety and availability. Technical competence and the autonomous action of German industry in the field of quality assurance set internationally recognized safety standards. The continuous exchange of experience through the active involvement of manufacturers, plant operators and third party inspection agencies in work establishing codes and standards and in th updating of the state of the art give the German system a control loop and feedback function (Technical Committees on Pressure Vessels). Within the framework of European harmonization it is a German concern that technical competence and expertise are not lost in a formally legal, bureaucratic certification procedure. In the course of the European harmonization process, the dual German QA concept should maintain its position by utilizing the specialist knowledge and competence of experts, and permit appropriate adaptation. (orig.)

Enhancement of submarine pressure hull steel ultrasonic inspection using imaging and artificial intelligence

Science.gov (United States)

Hay, D. Robert; Brassard, Michel; Matthews, James R.; Garneau, Stephane; Morchat, Richard

1995-06-01

The convergence of a number of contemporary technologies with increasing demands for improvements in inspection capabilities in maritime applications has created new opportunities for ultrasonic inspection. An automated ultrasonic inspection and data collection system APHIUS (automated pressure hull intelligent ultrasonic system), incorporates hardware and software developments to meet specific requirements for the maritime vessels, in particular, submarines in the Canadian Navy. Housed within a hardened portable computer chassis, instrumentation for digital ultrasonic data acquisition and transducer position measurement provide new capabilities that meet more demanding requirements for inspection of the aging submarine fleet. Digital data acquisition enables a number of new important capabilites including archiving of the complete inspection session, interpretation assistance through imaging, and automated interpretation using artificial intelligence methods. With this new reliable inspection system, in conjunction with a complementary study of the significance of real defect type and location, comprehensive new criteria can be generated which will eliminate unnecessary defect removal. As a consequence, cost savings will be realized through shortened submarine refit schedules.

Coupled finite difference and boundary element methods for fluid flow through a vessel with multibranches in tumours.

Science.gov (United States)

Sun, Qiang; Wu, Guo Xiong

2013-03-01

A mathematical model and a numerical solution procedure are developed to simulate flow field through a 3D permeable vessel with multibranches embedded in a solid tumour. The model is based on Poisseuille's law for the description of the flow through the vessels, Darcy's law for the fluid field inside the tumour interstitium, and Starling's law for the flux transmitted across the vascular walls. The solution procedure is based on a coupled method, in which the finite difference method is used for the flow in the vessels and the boundary element method is used for the flow in the tumour. When vessels meet each other at a junction, the pressure continuity and mass conservation are imposed at the junction. Three typical representative structures within the tumour vasculature, symmetrical dichotomous branching, asymmetrical bifurcation with uneven radius of daughter vessels and trifurcation, are investigated in detail as case studies. These results have demonstrated the features of tumour flow environment by the pressure distributions and flow velocity field. Copyright © 2012 John Wiley & Sons, Ltd.

Ultrasonic phased arrays for nondestructive inspection of forgings

International Nuclear Information System (INIS)

Wuestenberg, H.; Rotter, B.; Klanke, H.P.; Harbecke, D.

1993-01-01

Ultrasonic examinations on large forgings like rotor shafts for turbines or components for nuclear reactors are carried out at various manufacturing stages and during in-service inspections. During the manufacture, most of the inspections are carried out manually. Special in-service conditions, such as those at nuclear pressure vessels, have resulted in the development of mechanized scanning equipment. Ultrasonic probes have improved, and well-adapted sound fields and pulse shapes and based on special imaging procedures for the representation of the reportable reflectors have been applied. Since the geometry of many forgings requires the use of a multitude of angles for the inspections in-service and during manufacture, phased-array probes can be used successfully. The main advantages of the phased-array concept, e.g. the generation of a multitude of angles with the typical increase of redundancy in detection and quantitative evaluation and the possibility to produce pictures of defect situations, will be described in this contribution

Overview of the U.S. inspection capability demonstration program for CRDM penetrations

International Nuclear Information System (INIS)

Melton, M.; Whitaker, D.; Ammirato, F.; Pathania, R.

1994-01-01

The US PWR Owners have developed a coordinated program to demonstrate the effectiveness of procedures for inspection of the Control Rod Drive Mechanism (CRDM) nozzle penetrations. The purpose of the inspections is to detect inside surface Primary Water Stress Corrosion Cracking (PWSCC). Reactor vessel manufacturers, PWR Owners Groups, and EPRI are participating in this cooperative program. Flaw detection, sizing, and location capability will be assessed in the program using realistic full-scale mockups containing intentional defects. The program is expected to be completed by January 1, 1994 and be available for qualification work at that time. The demonstration program is scheduled to become operational in 1993 to support 1994 and future inspections

Device for inspecting the inside of a nuclear reactor

International Nuclear Information System (INIS)

Kurata, Masuo.

1985-01-01

Purpose: To apply illumination concentrically to a particular place and enable rapid and accurate inspection for the detailed portions by making such that the viewing direction of a television camera disposed in water and the direction of illumination of a submerged illumination tool can be varied optionally by remote control. Constitution: In the conventional inspection equipments for the inside of the reactor, a submerged illumination tool has merely been suspended by a cable and it has been difficult to effectively control the relative position with the television camera disposed in water. In this invention, the television camera disposed in water and a submerged illumination tool are rotatably disposed by way of a clamp to an inspection device main body. Then, the clamp can be rotated by the remote control and, further, the inspection device is suspended by way of a rope from a winding device disposed above the pressure vessel. In such a constitution, the submerged illumination tool can be adjusted to an optimal position to the television camera disposed in water. (Kamimura, M.)

Rochester’s Lead Law: Evaluation of a Local Environmental Health Policy Innovation

Science.gov (United States)

Ayoob, Maria; Morley, Rebecca

2011-01-01

Background: Significant progress has been made in reducing the incidence of childhood lead poisoning in the United States in the past three decades. However, the prevalence of elevated blood lead in children (≥ 10 μg/dL) remains high in some communities, particularly those with high proportions of pre-1978 housing in poor condition. Increasingly, municipalities are using local policy tools to reduce lead poisoning in high-risk areas, but little is known about the effectiveness of such policies. Objectives: In this article, we evaluated the effectiveness of a comprehensive rental housing–based lead law adopted in Rochester, New York, in 2005. Methods: This policy evaluation integrates analyses of city inspections data, a survey of landlords, landlord focus groups, and health department data on children’s blood lead levels from the first 4 years of implementation of the 2005 law. Results: Implementation has proceeded consistent with projected numbers of inspections with nearly all target units inspected in the first 4 years. Higher than expected inspection passage rates suggest that landlords have reduced lead hazards in rental housing affected by the law. Implementation of the lead law does not appear to have had a significant impact on the housing market. Conclusions: Although many uncertainties remain, our analysis suggests that the lead law has had a positive impact on children’s health. Strong enforcement, support for community-based lead programs, and ongoing intergovernmental coordination will be necessary to maintain lead-safe housing in Rochester. Lessons learned from the Rochester experience may inform future local lead poisoning prevention policies in other communities. PMID:22001644

Autonomous radiation monitoring of small vessels

International Nuclear Information System (INIS)

Ziock, K.P.; Cheriyadat, A.; Fabris, L.; Goddard, J.; Hornback, D.; Karnowski, T.; Kerekes, R.; Newby, J.

2011-01-01

Small private vessels are one avenue by which nuclear materials may be smuggled across international borders. While one can contemplate using the land-based approach of radiation portal monitors on the navigable waterways that lead to many ports, these systems are ill-suited to the problem. In contrast to roadways, where lanes segregate vehicles, and motion is well controlled by inspection booths; channels, inlets, and rivers present chaotic traffic patterns populated by vessels of all sizes. A unique solution to this problem is based on a portal-less portal monitor designed to handle free-flowing traffic on roadways with up to five-traffic lanes. The instrument uses a combination of visible-light and gamma-ray imaging to acquire and link radiation images to individual vehicles. This paper presents the results of a recent test of the system in a maritime setting.

Innovative inspection system for reactor pressure vessels; Innovative Pruefsysteme fuer Reaktordruckbehaelter

Energy Technology Data Exchange (ETDEWEB)

Mertens, K.; Trautmann, H.

1999-08-01

The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [Deutsch] Die vorgestellten kleinen, flexiblen und modernen Schwimmsysteme (DELPHIN-Manipulatoren und MIDAS-U-Boote) sind innovative Systeme fuer die Reduzierung der Aufwaende und Zeit zur Pruefung des Reaktordruckbehaelters und damit zur Reduktion der Revisionszeiten der Reaktoranlagen. (orig.)

Ultrasonic testing of electron beam closure weld on pressure vessel

International Nuclear Information System (INIS)

Andrews, R.W.

1975-01-01

One of the special products manufactured at the General Electric Neutron Devices Department (GEND) is a small stainless steel vessel designed to hold a component under high pressure for long periods. The vessel is a thick-walled cylinder with a threaded receptacle into which a plug is screwed and welded after receiving the unit to be tested. The test cavity is then pressurized through a small diameter opening in the bottom and that opening is welded closed. When x-ray inspection techniques did not reveal defective welds at the threaded plug in a pressured vessel, occasional ''leakers'' occurred. With normal equipment tolerances, the electron beam spike tends to wander from the desired path, particularly at the root of the weld. Ultrasonic techniques were used to successfully inspect the weld. The testing technique is based on the observation that ultrasonic energy is reflected from the unwelded screw threads and not from the regions where the threads are completely fused together by welding. Any gas pore or any threaded region outside the weld bead can produce an echo. The units are rotated while the ultrasonic transducer travels in a direction parallel to the axis of rotation and toward the welded end. This produces a helical scan which is converted to a two-dimensional presentation in which incomplete welds can be noted. (U.S.)

A development methodology for a remote inspection system with JAVA and socket

International Nuclear Information System (INIS)

Choi, Yoo Rark; Lee, Jae Cheol; Kim, Jae Hee

2004-01-01

We have developed RISYS (Reactor Inspection System) which inspects reactor vessel welds by an underwater mobile robot. The system consists of a main control computer and an inspection robot which is controlled by the main control computer. Since the environments of the inspection tasks in a nuclear plant, like in other industrial fields, is very poor, serious accidents often happen. Therefore the necessity for remote inspection and control system has increased more and more. We have carried out the research for a remote inspection model for RISYS, and have adopted the world wide web, java, and socket technologies for it. Client interface to access the main control computer that controls the inspection equipment is essential for the development of a remote inspection system. It has been developed with a traditional programming language, for example, Visual C++, Visual Basic and X-Window. However, it is too expensive to vend and maintain the version of a interface program because of the different computer O/S. Nevertheless web and java technologies come to the fore to solve the problems but the java interpreting typed language could incur a performance problem in operating the remote inspection system. We suggest a methodology for developing a remote inspection system with java, a traditional programming language, and a socket programming that solves the java performance problem in this paper

Law for controlling the use and applications of radioisotopes and ionizing radiations (Nuclear Law) Decreto Ley No. 11-86

International Nuclear Information System (INIS)

1986-01-01

This law defines the competent authority on radiation safety and includes all the requirements administrative, disposals on inspections, licensing, transport, records, authorizations and penalties about the use of radioisotopes and ionizing radiation in Guatemala

78 FR 30388 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel SAFARI; Invitation for Public...

Science.gov (United States)

2013-05-22

... Commercial Use Of Vessel: ``Sailing charters, tourism''. Geographic Region: ``Virginia, Maryland, Delaware... businesses in the U.S. that use U.S.-flag vessels. If MARAD determines, in accordance with 46 U.S.C. 12121... adverse effect on a U.S.-vessel builder or a business that uses U.S.-flag vessels in that business, a...

Inspection of Chooz power plant after ten years operation

International Nuclear Information System (INIS)

Saglio, Robert.

1978-01-01

This report is intended to discuss the results from the complete technical audit of the vessel effected in 1976 at the Ardennes reactor (CNA 305 MWe). This audit had a special character as far as this power plant has never been inspected before: the start-up had taken place in 1967 and was then prior to the development of French regulations (and even to the first version of the ASME code, Section XI). In that time, no inspection was expected; it has yet been possible to have a complete audit in ten days. The automatic start-up of focused ultrasonic testing so appeared to have reached the required reliability and a good sensitivity [fr

Navigation and vessel inspection circular No. 9-82. MSD certification. Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-05-10

This ciruclar is intended to provide information to the marine industry concerning MSD requirements and certification procedures. It is also intended to advise the Marine Industry that multiple certification of MSDs can and should be obtained depending upon the service of the vessel.

Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs

Energy Technology Data Exchange (ETDEWEB)

Khaleel, Mohammad A.; Simonen, Fredric A.

2009-05-01

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

Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

Energy Technology Data Exchange (ETDEWEB)

Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

Study on the quality of FRP fishing vessel based on improved Fishbone Chart

Science.gov (United States)

Sui, J. H.; Yu, Y. F.; Du, Q. F.; Jiang, D. W.

2018-01-01

The construction quality of FRP fishing vessels influences their production, use and industry development. In order to explore the factors that affect the construction quality of FRP fishing vessels, key factors affecting the construction quality of FRP fishing vessels are determined based on the quality problems of FRP fishing vessels constructed. The improved Fishbone Chart is used to analyze the eight factors of “human, machine, material, process, environment, inspection, design and information”. Taking the factors that affect the construction quality of FRP fishing vessels as the central target, the eight influencing factors were condensed into five aspects and a composite Fishbone Chart is drawn. The Fishbone Chart is used as the basic model, the influencing factors are sorted, screened and discriminated, and the system model convenient for construction site management and control is established. Finally, the causes of poor construction of FRP fishing vessels are analyzed and discussed, and relevant suggestions are put forward.

Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1989-10-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvement that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements based on material properties, service conditions, and NDE capabilities and uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1987 through March 1988. 21 refs., 28 figs., 2 tabs

Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1989-11-01

Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.

1993-09-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992

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)

The BWR vessel and internals project - 2001 and beyond

International Nuclear Information System (INIS)

Wagoner, V.; Mulford, T.

2001-01-01

The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

The BWR vessel and internals project - 2001 and beyond

Energy Technology Data Exchange (ETDEWEB)

Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

2001-07-01

The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

Fabrication of nuclear ship reactor MRX model and study on inspection and maintenance of components

International Nuclear Information System (INIS)

Kasahara, Yoshiyuki; Nakazawa, Toshio; Kusunoki, Tsuyoshi; Takahashi, Hiroki; Yoritsune, Tsutomu.

1997-10-01

The MRX (Marine Reactor X) is an integral type small reactor adopting passive safety systems. As for an integral type reactor, primary system components are installed in the reactor vessel. It is therefor important to establish the appropriate procedure for construction, inspection and maintenance, dismauntling, etc., for all components in the reactor vessel as well as in the reactor containment, because inspection space is limited. To study these subjects, a one-fifth model of the MRX was fabricated and operation capabilities were studied. As a result of studies, the following results are obtained. (1) Manufacturing and installing problems of the reactor pressure vessel, the containment vessel and internal components are basically not abserved. (2) Heat transfer tube structures of the steam generator and the heat exchangers of emergency decay heat removal system and containment water cooler were not seen of any problem for fabrication. However, due consideration is required in the detailed design of supports of heat transfer tubes. (3) Further studies should be needed for designs of flange penetrations and leak countermeasures for pipes instrument cables. (4) Arrangements of equipments in the containment should be taken in consideration in detail because the space is narrow. (5) Further discussion is required for installation methods of instruments and cables. (author)

Inspection qualification programme for VVER reactors and review of round robin test results

International Nuclear Information System (INIS)

Horacek, L.; Zdarek, J.

1998-01-01

Experience obtained, especially from in-service inspections of VVER 440-type reactor pressure vessels and from the Czech round test trials with international participation of ultrasonic teams, has highlighted the need for an in-service inspection qualification programme in the Czech Republic focused on NDT procedures, equipment and personnel. Recently, several national and international regional projects included in the PHARE programme (projects 4.1.2/93 and 1.02/94), briefly described, have been initiated. These projects are to cover step by step the programme of the in-service inspection qualification in view of technical justification as well as of practical assessment-performance demonstration-for all the main VVER-type primary circuit components. (orig.)

Progress and Achievements on the R&D Activities for ITER Vacuum Vessel

Energy Technology Data Exchange (ETDEWEB)

Nakahira, M. [Japan Atomic Energy Research Institute (JAERI); Koizumi, K. [Japan Atomic Energy Research Institute (JAERI); Takahashi, H. [Japan Atomic Energy Research Institute (JAERI); Onozuka, M. [ITER Joint Central Team, Garching, Germany; Ioki, K. [ITER Joint Central Team, Garching, Germany; Kuzumin, E. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Krylov, V. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Maslakowski, J. [Oak Ridge National Laboratory (ORNL); Nelson, Brad E [ORNL; Jones, L. [Max-Planck Institute, Garching, Germany; Danner, W. [Max-Planck Institute, Garching, Germany; Maisonnier, D. [Max-Planck Institute, Garching, Germany

2001-01-01

The ITER vacuum vessel (VV) is designed to be large double-walled structure with a D-shaped crosssection. The achievable fabrication tolerance of this structure was unknown due to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel, which was 15m in height, was fabricated and tested to obtain the fabrication and assembly tolerances. The model was fabricated within the target tolerance of 5mm and welding deformation during assembly operation was obtained. The port structure was also connected using remotized welding tools to demonstrate the basic maintenance activity. In parallel, the tests of advanced welding, cutting and inspection system were performed to improve the efficiency of fabrication and maintenance of the Vacuum Vessel. These activities show the feasibility of ITER Vacuum Vessel as feasible in a realistic way. This paper describes the major progress, achievement and latest status of the R&D activities on the ITER vacuum vessel.

77 FR 35744 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel ORIENTAL MYSTIQUE; Invitation...

Science.gov (United States)

2012-06-14

... Portland, Oregon and Oregon, City. The theme of these cruises will be on Chinese culture and nautical... businesses in the U.S. that use U.S.-flag vessels. If MARAD determines, in accordance with 46 U.S.C. 12121... adverse effect on a U.S.-vessel builder or a business that uses U.S.-flag vessels in that business, a...

Manufacture of EAST VS In-Vessel Coil

International Nuclear Information System (INIS)

Long, Feng; Wu, Yu; Du, Shijun; Jin, Huan; Yu, Min; Han, Qiyang; Wan, Jiansheng; Liu, Bin; Qiao, Jingchun; Liu, Xiaochuan; Li, Chang; Cai, Denggang; Tong, Yunhua

2013-01-01

Highlights: • ITER like Stainless Steel Mineral Insulation Conductor (SSMIC) used for EAST Tokamak VS In-Vessel Coil manufacture first time. • Research on SSMIC fabrication was introduced in detail. • Two sets totally four single-turn VS coils were manufactured and installed in place symmetrically above and below the mid-plane in the vacuum vessel of EAST. • The manufacture and inspection of the EAST VS coil especially the joint for the SSMIC connection was described in detail. • The insulation resistances of all the VS coils have no significant reduction after endurance test. -- Abstract: In the ongoing latest update round of EAST (Experimental Advanced Superconducting Tokamak), two sets of two single-turn Vertical Stabilization (VS) coils were manufactured and installed symmetrically above and below the mid-plane in the vacuum vessel of EAST. The Stainless Steel Mineral Insulated Conductor (SSMIC) developed for ITER In-Vessel Coils (IVCs) in Institute of Plasma Physics, Chinese Academy of Science (ASIPP) was used for the EAST VS coils manufacture. Each turn poloidal field VS coil includes three internal joints in the vacuum vessel. The middle joint connects two pieces of conductor which together form an R2.3 m arc segment inside the vacuum vessel. The other two joints connect the arc segment with the two feeders near the port along the toroidal direction to bear lower electromagnetic loads during operation. Main processes and tests include material performances checking, conductor fabrication, joint connection and testing, coil forming, insulation performances measurement were described herein

Study on in-vessel ISI for JOYO. Ultrasound propagation characteristic in the core support plate

International Nuclear Information System (INIS)

Ariyoshi, Masahiko; Ara, Kuniaki; Hirabayashi, Masaru

2005-03-01

The report describes the feasibility study on the in-vessel inspection technique to be applied for the experimental fast reactor JOYO. The object of this examination is to confirm the integration of reactor structure under sodium environment by an immediate means. The core support plate which is an important structure supports the weight of the core assembly is selected to an object of the inspection. In the examination until last year, the core support plate inspection equipment concept which combined ultrasound sensor with manipulator was constructed. In this concept, the ultrasound sensor is accessed to a low-pressure plenum sidewall and integrity of the core support plate weld is inspected. In this study, the ultrasound propagation behavior was examined to confirm the range where the core support plate by this concept was able to be inspected. The outline result is shown follows. (1) Only the transverse wave can be generated in the structure material by reflecting the incidence longitudinal wave from the sensor in the wedge. The use of this transverse wave is effective in the core support plate inspection. (2) Because the attenuation of the ultrasound wave depends on the distance, the sensor is made to approach from the fuel rack in the reactor vessel about two places in the upper part of the core support plate weld far from low-pressure plenum. (3) It is necessary to evaluate the permeability of the ultrasound wave by the mock-up examination in consideration of a peculiar attenuation of the structural material, the reflectivity from defect, etc. (4) In the core support plate inspection of phenix reactor, a weld about 4m away from the sensor position is inspected by using the Lamb wave. In this inspection, because it was generated to echo according to the geometrical shape of the structure material, the evaluation method by the analysis to identify the echo from the defect was constructed, and it was verified by the mock-up examination. It is preferable that

Analysis of inservice inspection relief requests

International Nuclear Information System (INIS)

Aldrich, D.A.; Cook, J.F.

1989-08-01

Nuclear Regulatory Commission (NRC) regulations require inspection (ISI) of boiling or pressurized water-cooled nuclear power plants be performed in accordance with a referenced edition and addenda of Section XI, ''Rules for Inservice Inspection of Nuclear Power Plant components,'' of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The regulations permit licensees to request relief from the NRC from specific ASME Code requirements that are determined to be impractical for the specific licensee. The NRC evaluates these requests and may grant such relief, but the NRC may also impose alternative or augmented inspections to assure structural reliability. The purpose,of this task was to evaluate the basis for ISI nondestructive examination (NDE) relief requests and to evaluate the effect of proposed ASME Code changes that would reduce the need for such requests or provide for more complete information in relief requests. This report contains the results of an analysis of an ISI relief request data base that has been expanded to include 1195 ISI relief requests versus the 296 relief requests covered in the first report in April 1987, EGG-SD-7430. Also relief requests were added to the data base which came from both first and second 10-year inspection intervals for several facilities. This provided the means to analyze the effect of recently approved ASME Code cases and updated Code requirements, some of which have been published as a result of earlier work on this task

Increasing for effectiveness of inspection of the use of radioactive substances well logging

International Nuclear Information System (INIS)

Hesty Rimadianny

2015-01-01

One of the utilization of nuclear power is the use of radioactive substances for well logging. To ensure the achievement of radiation safety and security in the use of radioactive substances for well logging activities BAPETEN carries out inspections in accordance with the mandate as sepulated in the Government Regulation No. 29 of 2008 on the Licensing Resource Utilization Ionizing Radiation and Nuclear Materials. Besides referring to the implementation of Government Regulation inspection also refers to BAPETEN Chairman Regulation No. 5 of 2009 on Radiation Safety in the Use of Radioactive material for well logging. In 2014, of 18 facilities inspected the most significant findings include the availability of equipment safety and security of radioactive substances, as well as the availability and suitability of documents and records of safety and security of radioactive substances for well logging activities. Based on these findings BAPETEN needs to make efforts to increase the effectiveness of inspections on the use of radioactive substances for well logging. Increasing the effectiveness of these inspections include a commitment for the frequency of well logging inspection, the number of qualified of inspectors in accordance with the established procedures and optimizing the law enforcement process which includes the application of administrative sanctions in the form of a written warning, license suspension, revocation until reporting to law enforcement. Besides, BAPETEN need to improve the effectiveness of outreach programs and legal guidance as a precaution in the long run. (author)

Acoustic emission test on a 25mm thick mild steel pressure vessel with inserted defects

International Nuclear Information System (INIS)

Bentley, P.G.; Dawson, D.G.; Hanley, D.J.; Kirby, N.

1976-12-01

Acoustic emission measurements have been taken on an experimental mild steel vessel with 4 inserted defects ranging in severity up to 90% of through thickness. The vessel was subjected to a series of pressure excursions of increasing magnitude until failure occurred by extension of the largest inserted defect through the vessel wall. No acoustic emission was detected throughout any part of the tests which would indicate the presence of such serious defects or of impending failure. Measurements of acoustic emission from metallurgical specimens are included and the results of post test inspection using conventional NDT and metallographic techniques are reported. (author)

Underwater television camera for monitoring inner side of pressure vessel

International Nuclear Information System (INIS)

Takayama, Kazuhiko.

1997-01-01

An underwater television support device equipped with a rotatable and vertically movable underwater television camera and an underwater television camera controlling device for monitoring images of the inside of the reactor core photographed by the underwater television camera to control the position of the underwater television camera and the underwater light are disposed on an upper lattice plate of a reactor pressure vessel. Both of them are electrically connected with each other by way of a cable to rapidly observe the inside of the reactor core by the underwater television camera. The reproducibility is extremely satisfactory by efficiently concentrating the position of the camera and image information upon inspection and observation. As a result, the steps for periodical inspection can be reduced to shorten the days for the periodical inspection. Since there is no requirement to withdraw fuel assemblies over a wide reactor core region, and the device can be used with the fuel assemblies being left as they are in the reactor, it is suitable for inspection of detectors for nuclear instrumentation. (N.H.)

Methodology if inspections to carry out the nuclear outages model; Metodologia de inspeccciones para cumplir el modelo de paradas nucleares

Energy Technology Data Exchange (ETDEWEB)

Aycart, J.; Mortenson, S.; Fourquet, J. M.

2005-07-01

Before the nuclear generation industry was deregulated in the United States, refueling and maintenance outages in nuclear power plants usually lasted orotund 100 days. After deregulation took effect, improved capability factors and performances became more important. As a result, it became essential to reduce the critical path time during the outage, which meant that activities that had typically been done in series had to be executed in parallel. The new outage model required the development of new tools and new processes, The 360-degree platform developed by GE Energy has made it possible to execute multiple activities in parallel. Various in-vessel visual inspection (IVVI) equipments can now simultaneously perform inspections on the pressurized reactor vessel (RPV) components. The larger number of inspection equipments in turn results in a larger volume of data, with the risk of increasing the time needed for examining them and postponing the end of the analysis phase, which is critical for the outage. To decrease data analysis times, the IVVI Digitalisation process has been development. With this process, the IVVI data are sent via a high-speed transmission line to a site outside the Plant called Center of Excellence (COE), where a team of Level III experts is in charge of analyzing them. The tools for the different product lines are being developed to interfere with each other as little as possible, thus minimizing the impact of the critical path on plant refueling activities. Methods are also being developed to increase the intervals between inspection. In accordance with the guidelines of the Boiling Water Reactor Vessel and Internals project (BWRVIP), the intervals between inspections are typically longer if ultrasound volumetric inspections are performed than if the scope is limited to IVVI. (Author)

Cracks on instrumentation penetrations in reactor vessel: a new challenge; Fissuration des penetrations de cuve: un nouveau defi

Energy Technology Data Exchange (ETDEWEB)

Anon

2004-02-01

In august 2003 NRC (nuclear regulatory commission) issued a warning concerning the deposits of boron acid that might accumulate on instrumental penetrations in the bottom of PWR vessels. These deposits were first detected on the David-Besse power plant and more recently on the unit 1 of South Texas Project (STP) during a refueling shutdown. STP contracted with the Areva company in order to perform inspections on all the 58 vessel penetrations of the unit 1 and to propose solutions. For that purpose the Areva company had to design a specific visual inspection tool that combined both ultra-sound method and Foucault current probing. The results of the inspection campaign on the unit 1 showed that only 2 penetration tubes were concerned with axial defects in their walls, that no circumferential defects were detected and that butt welds presented no cracks. The 2 incriminated penetration tubes were repaired: a section of both was replaced by an alloy-690 tube. (A.C.)

46 CFR 115.801 - Notice of inspection deficiencies and requirements.

Science.gov (United States)

2010-10-01

... the requirements of law or the regulations in this subchapter, the marine inspector will point out... presents a serious safety hazard to the vessel or it's passengers or crew, and exists through negligence or...

Assessment of inspectability of LMFBR designs. Final report

International Nuclear Information System (INIS)

1981-09-01

This two-volume report provides a comprehensive review of the inspectability of specific portions of loop- and pool-type LMFBR (1000-MWe) designs selected by EPRI. The designs were developed during the mid to late 1970s by three independent design teams (General Electric Co., Rockwell International, and Westinghouse) under the sponsorship of DOE (formerly ERDA) and EPRI. The requirements for normal, contingency, and post-repair inspections, addressed in this report, were established from Draft 12 of the ASME Boiler and Pressure Vessel Code, Section XI Division 3, issued in September 1979. These requirements, the intrinsic characteristics of the designs, the environmental (radiation, thermal, and atmospheric) aspects, and the available (present and near-term) inspection techniques, formed the basis for assessing the selected portions of the design or (1) accessibility, (2) feasibility, (3) practicality, and (4) costs to perform the above-specified inspections. Changes and additions fly ash has been as a concrete additive; however, extensive pilot scale development is underway to advance ash use in the TVA region in such areas as mineral and magnetite recovery, and mineral wool insulation. Recommended studies include: (1) the feasibility of converting existing wet fly d by the fuels include: residential (which includes residential and commercial), elthodology will be developed and verified in Phase II

In-vessel remote maintenance of the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

1987-01-01

The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

76 FR 77712 - Seagoing Barges

Science.gov (United States)

2011-12-14

... procedural and merely align the text of the regulations with current law and Coast Guard practice. An... submitted to our online docket via http://www.regulations.gov on or before February 13, 2012 or reaches the... Vessel Inspection Circular Pub. L. Public Law U.S.C. United States Code III. Regulatory Information We...

The photothermal camera - a new non destructive inspection tool

International Nuclear Information System (INIS)

Piriou, M.

2007-01-01

The Photothermal Camera, developed by the Non-Destructive Inspection Department at AREVA NP's Technical Center, is a device created to replace penetrant testing, a method whose drawbacks include environmental pollutants, industrial complexity and potential operator exposure. We have already seen how the Photothermal Camera can work alongside or instead of conventional surface inspection techniques such as penetrant, magnetic particle or eddy currents. With it, users can detect without any surface contact ligament defects or openings measuring just a few microns on rough oxidized, machined or welded metal parts. It also enables them to work on geometrically varied surfaces, hot parts or insulating (dielectric) materials without interference from the magnetic properties of the inspected part. The Photothermal Camera method has already been used for in situ inspections of tube/plate welds on an intermediate heat exchanger of the Phenix fast reactor. It also replaced the penetrant method for weld inspections on the ITER vacuum chamber, for weld crack detection on vessel head adapter J-welds, and for detecting cracks brought on by heat crazing. What sets this innovative method apart from others is its ability to operate at distances of up to two meters from the inspected part, as well as its remote control functionality at distances of up to 15 meters (or more via Ethernet), and its emissions-free environmental cleanliness. These make it a true alternative to penetrant testing, to the benefit of operator and environmental protection. (author) [fr

Quality assurance experience in the manufacture of PFBR reactor vessel during technology development work

International Nuclear Information System (INIS)

Shanmugam, K.; Chandramohan, R.; Ramamurthy, M.K.

1996-01-01

An efficient and proper implementation of quality assurance in the technology development works of Prototype Fast Breeder Reactor (PFBR) main vessel was undertaken to achieve the desired quality and dimensional accuracy of main vessel. In this paper an attempt has been made to bring out the methods and procedures adopted to implement the quality assurance programme on important activities including approval of documents, material, general requirements for manufacture of SS components, inspection procedures, forming and welding of petals, non-destructive testing etc. (author)

The IAEA Inspections: Creating a Collective Security System

International Nuclear Information System (INIS)

Le Guelte, Georges

2003-01-01

Through their successes, their achievements and the hopes they raised of a world where the law would replace the arbitrary, but also through the failures imputed to them and the fragility of their action, the inspections of the International Atomic Energy Agency (IAEA) have become a symbol of the capacities, the limits, and the weaknesses of the international system founded in 1945 by the UN Charter. Originally, their objective was much more limited. Historical developments - the Cold War, in particular - have turned the inspections into the major instrument of a unique mechanism of collective security, that is recovering its full significance in the current international context

Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

International Nuclear Information System (INIS)

Weiss, R.; Becker, R.; Lucht, B.; Mohr, F.; Hartwig, K.

2001-01-01

As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed. - Detection of surface connected longitudinal and transverse flaws: - notch, 3 mm deep and 10 mm long, for weave bead cladding; - notch, 2 mm deep and 20 mm long, for strip weld cladding. - Detection of embedded planar longitudinal and transverse flaws: - ligament of 7 mm for 8 mm clad thickness and 3 mm; - ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm. - Detection of embedded volumetric longitudinal and transverse flaws: - 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep

A study on detection of internal defects of pressure vessel by digital shearography

International Nuclear Information System (INIS)

Kang, Young Jun; Park, Sung Tae; Lee, Hae Moo; Nam, Seung Hun

1999-01-01

Pipelines in power plants, nuclear facilities and chemical industries are often affected by corrosion effects. The inspection of internal defects of these pipelines is important to guarantee safe operational condition. Conventional NDT methods have been taken relatively much time, money, and manpower because of performing as the method of contact with objects to be inspected. Digital shearography is a laser-based optical method which allows full-field observation of surface displacement derivatives. This method has many advantages in practical use, such as low sensitivity to environmental noise, simple optical configuration and real time measurement. Therefore it is a good method to use for detecting internal defects. In this paper, the experiment was performed with some pressure vessels which has different internal cracks. We detected internal cracks of the pressure vessels at a real time and evaluated qualitatively these results. We also performed qualitative measurement of shearo fringe by using phase shifting method.

Effectiveness and reliability of US inservice inspection techniques

International Nuclear Information System (INIS)

Doctor, S.R.; Becker, F.L.; Selby, G.P.

1982-01-01

The work presented is from an ongoing program directed toward measuring the effectiveness and reliability of inservice inspection (ISI) of light water reactor systems (primary piping and pressure vessel). Extensive round robin and parametric evaluations have been conducted in 10 in. Sch. 80 stainless steel as well as centrifugally cast stainless steel and clad ferritic main coolant pipe welds. Results from these measurements will be viewed in relationship to US regulations and ASME Section XI Code requirements. 6 figures

Experience in surveillance of the prestress of concrete reactor vessels in Wylfa nuclear power station

International Nuclear Information System (INIS)

Dawson, P.; Paton, A.A.; Walsh, S.R.

1989-01-01

This paper describes experience gained in the in-service surveillance of the prestressing system for the prestressed concrete reactor vessels (PCRVs) at Wylfa nuclear power station. The paper gives details of results for the prestressing system obtained from the statutory in-service inspection program of the PCRVs. The program includes a detailed examination of a selection of prestressing tendon anchorages, anchorage load checks using a lift-off technique on a one percent sample of tendons and corrosion inspection of samples of prestressing strand and determination of their mechanical properties. The results obtained from the above in-service inspections have shown that the prestressing system continues to function within its design limits

Application of phased arrays in basic and in-service inspection

International Nuclear Information System (INIS)

Gebhardt, W.; Schwarz, H.P.; Bonitz, F.; Woll, H.

1985-01-01

In the scope of the reactor safety research program of the Federal Ministry of Research and Technology a flexible microcomputer controlled phased array system was developed. Meanwhile, several industrial prototypes for simple and complicated applications are built up. The applicability of phased array systems in NDE for basic and inservice inspections of reactor pressure vessels is investigated. Methods for defect detection, reconstruction and classification are described

Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas

Energy Technology Data Exchange (ETDEWEB)

Alberdi, R.; Gonzalez, E.; Willke, A.; Yague, L. [TECNATOM SA, Madrid (Spain)

2001-07-01

In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. The decision was taken to perform ID inspections, mainly because this type of scanners were not available at that time, and which provide the 4 following advantages. 1) Maximum inspected weld length. This avoids interference with the jet pumps and the systems present in the annulus and affecting OD inspections. Besides, the repairs performed on in-service core shrouds in all cases imply the addition of new fixed elements on their outer surface, since the fuel assembly space must be left free. 2) Reduction of inspection times and of unforeseen events: maintenance of planning schedules, reduction of personnel doses, reduced critical path time. 3) High inspection accuracy and repeatability. 4) Simplification of equipment positioning work (similar to the installation of fuel assemblies). As regards inspection techniques, the

''Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas''

Energy Technology Data Exchange (ETDEWEB)

Alberdi, R.; Willke, A.; Gonzalez, E.; Yague, L

2001-07-01

In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. As regards inspection techniques, the decision was taken to carry out acquisition simultaneously using both ultrasonics (UT) and eddy currents (ET). (author)