The ASME research task force on risk-based in-service inspection

International Nuclear Information System (INIS)

Balkey, K.R.; Chapman, O.J.V.

1997-01-01

The use of risk-based methods in the development of in-service inspection (ISI) and in-service testing (IST) programs for nuclear power plant and other industrial applications has been studied for the last several years through the American Society of Mechanical Engineers Centre for Research and Technology Development (ASME 1991, 1992, 1994, 1996). The results of this work are being used as a foundation to develop specific requirements for implementation of risk-based technology in ASME Codes and Standards, regulatory requirements and industry programs both in the U.S. and other countries. This paper provides a brief overview of the ASME Research Methodology and how it has been adapted for application to the inspection of piping within the USA. It also relates how the reliability of nondestructive examination (NDE) methods for pressure boundary components can impact the risk and discusses the relationship between this and NDE qualification/demonstration now being implemented in Europe and the USA. (orig.)

Strategy of Risk-Informed Inspection for Secondary Systems in NPPs

International Nuclear Information System (INIS)

Kim, So Young; Heo, Gyun Young

2012-01-01

Recently, unexpected accidents such as large-scale blackout have been occurred due to increasing demand of electricity. The unplanned trips have great influence over the economics although they rarely affect the safety of system. Because the cost for inspection and maintenance is limited, it is beneficial not only economically to take follow-up action in a timely manner to prevent plant trip but also in safety by protecting against the risk which is occurred from trip. In nuclear power plants (NPPs), the importance about risk-informed service such as inspection, maintenance and replacement is becoming on the rise. Risk-informed techniques such as in-service test, in-service inspection, design guide change and etc. are already used in nuclear field but these techniques are applied focusing on primary system for safety-related regulatory issues. It is needed to consider the optimization of inspection informed risk for entire plant including secondary system which is not applied enough. The study about risk-based inspection for boilers in thermal power field is in progress actively, but not for turbine cycle, which is similar with nuclear industry. Today, the most of inspections or tests for turbine cycle are time-based maintenance, being performed on time. It is required to perform condition-based maintenance by evaluating risk analyzing both of probability and consequence simultaneously about damage mechanisms that can be predicted by equipment or devices configuring the secondary system. Probability of failure (POF) is analyzed in terms of materials' ageing mechanisms, but consequence of failure (COF) should be evaluated in terms of thermohydraulic condition then it is needed to check power loss and period of maintenance. This paper is going to propose the idea to quantify the consequence of failure using the fault tree and simulation for secondary system, and propose quantitative risk-informed inspection by these approaches

In-service inspection techniques

International Nuclear Information System (INIS)

Backfisch, W.; Zipser, R.R.

1980-01-01

The owner of a nuclear power plant (NPP) is obligated and interested to maintain - by regular maintenance and in-service inspections - the operational safety and availability of the plant for the subsequent operating period in a condition, as is specified as the basis of the erection and the last operational permits. In-service inspections are performed to verify the operational safety, and maintenance work is performed to guarantee the availability. Below, the typical in-service inspections of a light-water reactor NPP (operated on a pressurized-water reactor or on a boiling-water reactor) are described with details and examples of typical inspections, especially of recurrent performance tests of the systems. (orig./RW)

Selection of procedures for inservice inspections; Auswahl der Verfahren fuer wiederkehrende Pruefungen

Energy Technology Data Exchange (ETDEWEB)

Brast, G [Preussische Elektrizitaets-AG (Preussenelektra), Hannover (Germany); Britz, A [Bayernwerk AG, Muenchen (Germany); Maier, H J [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Seidenkranz, T [TUEV Energie- und Systemtechnik GmbH, Mannheim (Germany)

1998-11-01

At present, selection of procedures for inservice inspection has to take into account the legal basis, i.e. the existing regulatory codes, and the practical aspects, i.e. experience and information obtained by the general, initial inservice inspection or performance data obtained by the latest, recurrent inspection. However, regulatory codes are being reviewed to a certain extent in order to permit integration of technological progress. Depending on the degree of availability in future, of inspection task-specific, sensitive and qualified NDE techniques for inservice inspections (`risk based ISI`), the framework of defined inspection intervals, sites, and detection limits will be broken up and altered in response to progress made. This opens up new opportunities for an optimization of inservice inspections for proof of component integrity. (orig./CB) [Deutsch] Zur Zeit muss sich die Auswahl der Pruefverfahren an den gueltigen Regelwerken und, da es sich um wiederkehrende Pruefungen handelt, an der Basispruefung bzw. der letzten wiederkehrenden Pruefung orientieren. Jedoch vollzieht sich zur Zeit eine Oeffnung der Regelwerke, mit der man auch der Weiterentwicklung der Prueftechniken Rechnung traegt. In dem Masse, wie zukuenftig auf die Pruefaufgabe/Pruefaussage optimal abgestimmte und qualifizierte Prueftechniken mit einer hohen Nachweisempfindlichkeit am Bauteil fuer zielgerichtete wiederkehrende Pruefungen (als `risk based ISI`) zur Verfuegung stehen, wird der Rahmen mit festgelegten Pruefintervallen, Prueforten und festen Registriergrenzen gesprengt und variabel gestaltet werden koennen. Damit ergeben sich neue Moeglichkeiten fuer eine Optimierung der WKP zum Nachweis der Integritaet des Bauteils. (orig./MM)

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.

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.

A pilot application of risk-informed methods to establish inservice inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station. Revision 1

International Nuclear Information System (INIS)

Vo, T.V.; Phan, H.K.; Gore, B.F.; Simonen, F.A.; Doctor, S.R.

1997-02-01

As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest National Laboratory has developed risk-informed approaches for inservice inspection plans of nuclear power plants. This method uses probabilistic risk assessment (PRA) results to identify and prioritize the most risk-important components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot application of this methodology. This report, which incorporates more recent plant-specific information and improved risk-informed methodology and tools, is Revision 1 of the earlier report (NUREG/CR-6181). The methodology discussed in the original report is no longer current and a preferred methodology is presented in this Revision. This report, NUREG/CR-6181, Rev. 1, therefore supersedes the earlier NUREG/CR-6181 published in August 1994. The specific systems addressed in this report are the auxiliary feedwater, the low-pressure injection, and the reactor coolant systems. The results provide a risk-informed ranking of components within these systems

Lessons Learned From Implementation of Westinghouse Owners Group Risk-Informed Inservice Inspection Methodology for Piping

International Nuclear Information System (INIS)

Stevenson, Paul R.; Haessler, Richard L.; McNeill, Alex; Pyne, Mark A.; West, Raymond A.

2006-01-01

Risk-informed inservice inspection (ISI) programs have been in use for over seven years as an alternative to current regulatory requirements in the development and implementation of ISI programs for nuclear plant piping systems. Programs using the Westinghouse Owners Group (WOG) (now known as the Pressurized Water Reactor Owners Group - PWROG) risk-informed ISI methodology have been developed and implemented within the U.S. and several other countries. Additionally, many plants have conducted or are in the process of conducting updates to their risk-informed ISI programs. In the development and implementation of these risk-informed ISI programs and the associated updates to those programs, the following important lessons learned have been identified and are addressed. Concepts such as 'loss of inventory', which are typically not modeled in a plant's probabilistic risk assessment (PRA) model for all systems. The importance of considering operator actions in the identification of consequences associated with a piping failure and the categorization of segments as high safety significant (HSS) or low safety significant (LSS). The impact that the above considerations have had on the large early release frequency (LERF) and categorization of segments as HSS or LSS. The importance of automation. Making the update process more efficient to reduce costs associated with maintaining the risk-informed ISI program. The insights gained are associated with many of the steps in the risk-informed ISI process including: development of the consequences associated with piping failures, categorization of segments, structural element selection and program updates. Many of these lessons learned have impacted the results of the risk-informed ISI programs and have impacted the updates to those programs. This paper summarizes the lessons learned and insights gained from the application of the WOG risk-informed ISI methodology in the U.S., Europe and Asia. (authors)

Challenges in using a probabilistic safety assessment in a risk informed process (illustrated using risk informed inservice inspection)

International Nuclear Information System (INIS)

Chapman, James R.; Dimitrijevic, Vesna B.

1999-01-01

Many of the ongoing and expected uses of Probabilistic Safety Assessment (PSA) create new challenges to ensuring that the resulting conclusions are valid. This paper provides a summary of some of these challenges. Work conducted by the authors on Risk-Informed Inservice Inspection (RI-ISI) is used to illustrate these challenges. Means to address all of the challenges are not provided in detail in this paper. Several earlier papers discuss how these challenges can be addressed. References are provided for the interested reader (Chapman JR et al. In: PSA '95, vol. 1, Seoul, 1995: 177-80; Chapman JR et al. In: ICONE-IV, New Orleans, 1996; Dimitrijevic VB et al. In: Croatian Nuclear Society International Conference, Opatija, 1996: 245-54; Dimitrijevic VB et al. In: Croatian Nuclear Society International Conference, Opatija, 1996: 255-62; Dimitrijevic VB. In: Yugoslav Nuclear Society Conference, Belgrade, 1996: 53-61; O'Regan PJ et al. In: PSA '95, Seoul, vol. 1, 1995: 403-5; O'Regan PJ. In: ICONE-IV, vol. 5, New Orleans, 1996: 277-80)

In-service inspections of V-230 reactor

International Nuclear Information System (INIS)

Prepechal, J.

1984-01-01

It is stated that despite certain constraints the configuration of the WWER-440 is such that it allows to make in-service inspections on a fully satisfactory scale. Three factors are discussed whose existence is necessary for the implementation of in-service inspections. The program defining the scale of inspections is satisfactory with regard to the safety and reliability of reactor operation. Its further development must result in reducing time consumption and radiation burden of personnel. Regulations for the implementation and evaluation of inspections represent the weakest link in the system of in-service inspections. At present, various organizations are dealing with the said problem within international cooperation. Equipment for in-service inspections of WWER-440 reactors is relatively good. The most important knowledge is summed up gained from the ten pre-service and in-service inspections of reactors of this type made so far. (Z.M.)

Risk-Informed Decisions Optimization in Inspection and Maintenance

International Nuclear Information System (INIS)

Robertas Alzbutas

2002-01-01

The Risk-Informed Approach (RIA) used to support decisions related to inspection and maintenance program is considered. The use of risk-informed methods can help focus the adequate in-service inspections and control on the more important locations of complex dynamic systems. The focus is set on the highest risk measured as conditional core damage frequency, which is produced by the frequencies of degradation and final failure at different locations combined with the conditional failure consequence probability. The probabilities of different degradation states per year and consequences are estimated quantitatively. The investigation of inspection and maintenance process is presented as the combination of deterministic and probabilistic analysis based on general risk-informed model, which includes the inspection and maintenance program features. Such RIA allows an optimization of inspection program while maintaining probabilistic and fundamental deterministic safety requirements. The failure statistics analysis is used as well as the evaluation of reliability of inspections. The assumptions regarding the effectiveness of the inspection methods are based on a classification of the accessibility of the welds during the inspection and on the different techniques used for inspection. The probability of defect detection is assumed to depend on the parameters either through logarithmic or logit transformation. As example the modeling of the pipe systems inspection process is analyzed. The means to reduce a number of inspection sites and the cumulative radiation exposure to the NPP inspection personnel with a reduction of overall risk is presented together with used and developed software. The developed software can perform and administrate all the risk evaluations and ensure the possibilities to compare different options and perform sensitivity analysis. The approaches to define an acceptable level of risk are discussed. These approaches with appropriate software in

Status report on developments and cooperation on risk-informed inservice-inspection and non-destructive testing (NDT) qualification in OECD-NEA member countries - CSNI integrity and ageing working group

International Nuclear Information System (INIS)

Skanberg, Lars

2005-01-01

During the last decades the nuclear industry has experienced service degradation of many components, both in primary systems and in secondary systems. Several inspection failures have also occurred over the years. This degradation and inspection history as well as the economical and political factors have consequently set up a pressure for more efficient and cost-effective in-service inspection programmes to ensure that there are adequate safety margins so that anticipated degradations of components do not lead to failures that cause accidents or even unplanned shutdowns with adverse effects on power production reliability. In this situation, nuclear regulators as well as nuclear utilities in many countries have developed and implement risk-informed inspection approaches together with more stringent requirements of demonstrating the performance of the NDT systems that are to be used for inspection of safety related components which are susceptible to different kind of degradation mechanisms. In December 2000, the Committee on Nuclear Regulatory Activities (CNRA) and the Committee on the Safety of Nuclear Installations (CSNI) agreed to prepare a state-of-the art report addressing the present situation and regulatory aspects in NEA member countries on: - Risk based / risk informed in-service inspections developments, - Qualification of NDT system to be used for the inspections. The CSNI gave mandate to the CSNI working group on the Integrity of Components and Structures (IAGE) to prepare the report. Practices and status in NEA member countries were collected in 2003 through a questionnaire. Results have been compiled in the report NEA/CSNI/R(2005)3. To complete the technical information, a CSNI Workshop was held from 13 to 14 April 2004 in Stockholm, Sweden hosted by the Swedish Nuclear Power Inspectorate. The Workshop gathered 54 participants from 17 countries including the EC, the IAEA and the main organisations worldwide developing RI-ISI methodologies. Papers

Feasibility of developing risk-based rankings of pressure boundary systems for inservice inspection

Energy Technology Data Exchange (ETDEWEB)

Vo, T.V.; Smith, B.W.; Simonen, F.A.; Gore, B.F.

1994-08-01

The goals of the Evaluation and Improvement of Non-destructive Examination Reliability for the In-service Inspection of Light Water Reactors Program sponsored by the Nuclear Regulatory Commission at Pacific Northwest Laboratory (PNL) are to (1) assess current ISI techniques and requirements for all pressure boundary systems and components, (2) determine if improvements to the requirements are needed, and (3) if necessary, develop recommendations for revising the applicable ASME Codes and regulatory requirements. In evaluating approaches that could be used to provide a technical basis for improved inservice inspection plans, PNL has developed and applied a method that uses results of probabilistic risk assessment (PRA) to establish piping system ISI requirements. In the PNL program, the feasibility of generic ISI requirements is being addressed in two phases. Phase I involves identifying and prioritizing the systems most relevant to plant safety. The results of these evaluations will be later consolidated into requirements for comprehensive inservice inspection of nuclear power plant components that will be developed in Phase II. This report presents Phase I evaluations for eight selected plants and attempts to compare these PRA-based inspection priorities with current ASME Section XI requirements for Class 1, 2 and 3 systems. These results show that there are generic insights that can be extrapolated from the selected plants to specific classes of light water reactors.

Feasibility of developing risk-based rankings of pressure boundary systems for inservice inspection

International Nuclear Information System (INIS)

Vo, T.V.; Smith, B.W.; Simonen, F.A.; Gore, B.F.

1994-08-01

The goals of the Evaluation and Improvement of Non-destructive Examination Reliability for the In-service Inspection of Light Water Reactors Program sponsored by the Nuclear Regulatory Commission at Pacific Northwest Laboratory (PNL) are to (1) assess current ISI techniques and requirements for all pressure boundary systems and components, (2) determine if improvements to the requirements are needed, and (3) if necessary, develop recommendations for revising the applicable ASME Codes and regulatory requirements. In evaluating approaches that could be used to provide a technical basis for improved inservice inspection plans, PNL has developed and applied a method that uses results of probabilistic risk assessment (PRA) to establish piping system ISI requirements. In the PNL program, the feasibility of generic ISI requirements is being addressed in two phases. Phase I involves identifying and prioritizing the systems most relevant to plant safety. The results of these evaluations will be later consolidated into requirements for comprehensive inservice inspection of nuclear power plant components that will be developed in Phase II. This report presents Phase I evaluations for eight selected plants and attempts to compare these PRA-based inspection priorities with current ASME Section XI requirements for Class 1, 2 and 3 systems. These results show that there are generic insights that can be extrapolated from the selected plants to specific classes of light water reactors

Nuclear power plant pressure vessels. Inservice inspections

International Nuclear Information System (INIS)

1995-01-01

The requirements for the planning and reporting of inservice inspections of nuclear power plant pressure vessels are presented. The guide specifically applies to inservice inspections of Safety class 1 and 2 nuclear power plant pressure vessels, piping, pumps and valves plus their supports and reactor pressure vessel internals by non- destructive examination methods (NDE). Inservice inspections according to the Pressure Vessel Degree (549/73) are discussed separately in the guide YVL 3.0. (4 refs.)

In-service inspection and periodic testing

International Nuclear Information System (INIS)

Eisele, H.; Meyer, F.A.; Zipser, R.R.

1981-01-01

In-service inspections are performed to verify the operational safety, and maintenance work is performed to guarantee the availability. Below, the typical in-service inspections of a light-water reactor NPP (operated on a pressurized-water reactor/PWR/ or on a boiling-water reactor/BWR/) are described with details and examples of typical inspections, especially of recurrent performance tests of the systems. (orig./RW)

In-service inspection and periodic testing

International Nuclear Information System (INIS)

Eisele, H.; Meyer, F.A.; Zipser, R.R.

1980-01-01

In-service inspections are performed to verify the operational safety, and maintenance work is performed to guarantee the availability. In the present paper, the typical in-service inspections of a light-water reactor NPP (operated on a pressurized-water reactor/PWR/ or on a boiling-water reactor/BWR/) are described with details and examples of typical inspections, especially of recurrent performance tests of the systems. (orig./RW)

Development of a risk-based in-service inspection program for an LMR

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.

1996-01-01

The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. Experimental Breeder Reactor II (EBR-II) is a liquid-metal-cooled fast reactor (LMR) that operated for 30 yr before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the LMR design. For example, it demonstrated passive response to plant upset events, low-pressure primary coolant, and compatibility of the coolant and reactor materials. This ISI program was based on work currently being done by an American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection

Risk-based inspection--Development of guidelines

International Nuclear Information System (INIS)

1993-07-01

Effective inservice inspection programs can play a significant role in minimizing equipment and structural failures. Most of the current inservice inspection programs for light water reactor (LWR) nuclear power plant components are based on experience and engineers' qualitative judgment. These programs include only an implicit consideration of risk, which combines the probability of failure of a component under its operation and loading conditions and the consequences of such failure, if it occurs. This document recommends appropriate methods for establishing a risk-based inspection program for LWR nuclear power plant components. The process has been built from a general methodology (Volume 1) and has been expanded to involve five major steps: defining the system; evaluating qualitative risk assessment results; using this and information from plant probabilistic risk assessments to perform a quantitative risk analysis; selecting target failure probabilities; and developing an inspection program for components using economic decision analysis and structural reliability assessment methods

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

Development and testing of VTT approach to risk-informed in-service inspection methodology. Final report of SAFIR INTELI INPUT Project RI-ISI

International Nuclear Information System (INIS)

Cronvall, O.; Maennistoe, I.; Simola, K.

2007-04-01

This report summarises the results of a research project on risk-informed in-service inspection (RI-ISI) methodology conducted in the Finnish national nuclear energy research programme SAFIR (2003-2006). The purpose of this work was to increase the accuracy of risk estimates used in RI-ISI analyses of nuclear power plant (NPP) piping systems, and to quantitatively evaluate the effects of different piping inspection strategies on risk. Piping failure occurrences were sampled by using probabilistic fracture mechanics (PFM) analyses. The PFM results for crack growth were used to construct transition matrices for a discrete-time Markov process model, which in turn was applied to examine the effects of various inspection strategies on the failure probabilities and risks. The applicability of the developed quantitative risk matrix approach was examined as a pilot study performed to the Shut-down cooling piping system 321 in NPP unit OL1 of Teollisuuden Voima Oy (TVO). The analysed degradation mechanisms were stress corrosion cracking (SCC) and thermal fatigue induced cracking (in the mixing points). Here a new and rather straightforward approach was developed to model thermal fatigue induced cracking, which degradation mechanism is much more difficult to model than SCC. This study further demonstrated the usefulness of Markov analysis procedure development by VTT in RI-ISI applications. The most important results are the quantified comparisons of different inspections strategies. It was shown in this study that Markov models are useful for this purpose, when combined with PFM analyses. While the numerical results could benefit from further considerations of inspection reliability, this does not affect the feasibility of the method itself. The approach can be used to identify an optimal inspection strategy for achieving a balanced risk profile of piping segments. (orig.)

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

In-service inspection methods for graphite-epoxy structures on commercial transport aircraft

Science.gov (United States)

Phelps, M. L.

1981-01-01

In-service inspection methods for graphite-epoxy composite structures on commercial transport aircraft are determined. Graphite/epoxy structures, service incurred defects, current inspection practices and concerns of the airline and manufacturers, and other related information were determind by survey. Based on this information, applicable inspection nondestructive inspection methods are evaluated and inspection techniques determined. Technology is developed primarily in eddy current inspection.

Planning and programming of pre-operational and in-service inspections

International Nuclear Information System (INIS)

Udaondo, M.S.

1984-01-01

After a brief mention of the actual scope of in-service inspection work, conclusions are presented that relate to the preparations for inspections, making use of experience acquired since 1972 at 11 nuclear power plants in Spain which have commissioned such studies from one particular organization, and analyses are given of the advantages to be gained therefrom. Three different aspects of preparations for inspections are considered: (a) man-hour estimates and the duration of in-service inspections; (b) the sequence of action during a pre-operational inspection (assuming a typical functional organization) from definition of the codes of practice and standards applicable up to the issue of the final report and the schedule for distribution of the annual work load to be invested in a typical project, as a result of combining the two previous estimates, and (c) the documentary aspect of preparations for an inspection during a scheduled outage, as related to the various documents to be drawn up and their contents. Reference is made to the general training to be given to the staff in charge of inspection activities so as to provide them with information on, and a perspective of, the in-service inspection jobs required. (author)

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

In-service inspection guidelines for composite aerospace structures

International Nuclear Information System (INIS)

Heida, Jaap H.; Platenkamp, Derk J.

2012-01-01

The in-service inspection of composite aerospace structures is reviewed, using the results of a evaluation of promising, mobile non-destructive inspection (NDI) methods. The evaluation made use of carbon fibre reinforced specimens representative for primary composite aerospace structures, including relevant damage types such as impact damage, delaminations and disbonds. A range of NDI methods were evaluated such as visual inspection, vibration analysis, phased array ultrasonic inspection, shearography and thermography inspection. Important aspects of the evaluation were the capability for defect detection and characterization, portability of equipment, field of view, couplant requirements, speed of inspection, level of training required and the cost of equipment. The paper reviews the damage tolerance design approach for composites, and concludes with guidelines for the in-service inspection of composite aerospace structures.

Non-destructive inservice inspections

International Nuclear Information System (INIS)

Kauppinen, P.; Sarkimo, M.; Lahdenperae, K.

1998-01-01

In order to assess the possible damages occurring in the components and structures of operating nuclear power plants during service the main components and structures are periodically inspected by non-destructive testing techniques. The reliability of non-destructive testing techniques applied in these inservice inspections is of major importance because the decisions concerning the needs for repair of components are mainly based on the results of inspections. One of the targets of this research program has been to improve the reliability of non-destructive testing. This has been addressed in the sub-projects which are briefly summarised here. (author)

Development of a risk-based inservice inspection program for a liquid metal reactor

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.

1996-01-01

The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. EBR-II is liquid-metal-cooled fast reactor which operated for thirty years before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the liquid-metal reactor design, e.g., demonstrated passive response to plant upset events, low pressure primary coolant and compatibility of the coolant and reactor materials. Many of the systems cannot be inspected due to inaccessibility of the components. However, application of a risk-based approach provided the basis for reducing or eliminating inspections in some areas that would otherwise be required. Development and implementation of the risk-based ISI program was interrupted by the DOE-mandated shutdown of EBR-II, so the potential benefits of this approach in terms of reduced O and M costs have yet to be realized. Through the development of this program, however it is clear that there is potential for substantial cost-savings while improving the risk-profile of the facility through this approach

Application of risk based inspection as a part of life management of nuclear power plants

International Nuclear Information System (INIS)

Vinod, Gopika; Babar, A.K.; Saraf, R.K.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Full text: Risk Informed approach is a systematic effort to improve plant safety in a more efficient manner by distributing the available safety resources depending on the importance to plant safety. This approach has found immense application in various aspects associated with Nuclear Power Plants, including design, manufacturing, operation and regulation. Typical applications of Risk Informed approach are in Technical Specification, In-Service Inspection (ISI) requirements, Motor Operated Valve testing, Configuration Control etc. In order to cater to such variety of Probabilistic Safety Assessment (PSA) based Risk Informed requirements, a software package, 'Risk Monitor' has been developed by the authors. An important application of Risk informed approach that has been undertaken for Indian Pressurised Heavy Water Reactor (IPHWR) is towards Risk Informed In-Service Inspection (RI-ISI). Studies are being conducted to evolve an inspection plan that is optimised to provide effective inspections at the right location with a proper inspection frequency. Using risk informed approach, the identification of system / component for inspection resource allocation is based on the results from Level 1 PSA of a NPP. Even though this methodology can be employed on any of the nuclear components/systems such as mechanical systems, instrumentation, etc., as a starting point, piping has been considered for employing Risk Informed Inspection. Plant risk assessment is modelled through the analysis of Core Damage Frequency (CDF) using PSA models. Risk Informed Inspection program of piping involves the estimation of failure probability or frequency of a piping segment and estimation of consequences of piping failures. ISI program changes could affect the failure probability values of piping and can introduce a change in CDF. These effects can be brought out through the implementation of Risk Informed inspection strategy. Various importance measures like Fussel-Vesely, Birnbaum

Review of activities relevant to in-service inspection

International Nuclear Information System (INIS)

Imanaka, N.

1980-01-01

Nuclear power plants are requested to provide continuing safety that cannot compare with other industries, as plant safety is a matter of much concern. To provide continuous assurance for plant safety there is increasing tendency to demand much of inspection of components during the lifetime. This inservice inspection of LMFBRs should be investigated from a view point of different systems and characteristics from LWRs. In this paper a review for inservice inspection of LMFBRs is described. To provide a continuous assurance of safety to the LMFBR, it is essential to develop how to construct the components to maintain the integrity throughout the service lifetime. Especially how to design is urged for this object. In-service inspection should be located only to compensate some uncertainty remained at the design stage, as it is too much complex in practice. As for inspection techniques, leak monitoring is assumed to be a best way to assure the plant safety continuously with the minimum plant outage time and minimum radioactive hazard to the inspectors

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)

New developments in containment in-service inspection requirements

International Nuclear Information System (INIS)

Staffiera, J.E.

1995-01-01

Section 11 of the ASME Boiler and Pressure Vessel Code contains requirements for inservice inspection of nuclear power plant components. Development of ASME Code requirements for containment inservice inspection was begun in 1977, and in 1979 the first such requirements were published in the form of Code Case N-236. Formal inclusion of these requirements in Section 11 of the ASME Code occurred with publication of Subsection IWE, ''Rules for inservice Inspection of Class MC Components of Nuclear Power Plants,'' in the 1980 Edition, Winter 1981 Addenda. At that time, inspection emphasis on nuclear power construction and operation activities was placed on welds and welding processes associated with steel containments and metallic liners of concrete containments. The need for repair-welding requirements was necessitated by containment design modifications for conditions not considered in several original plant designs. Welds in steel containments and metallic liners of concrete containments have not required significant amounts of repair, however, degradation of base metal in containments has become a major concern. Various degradation mechanisms have been identified as potential causes of damage to containment surfaces, including fatigue, corrosion and material embrittlement due to long-term radiation exposure. As a result of these concerns, and in response to comments generated by the Committee to Review Generic Requirements (CRGR) of the NRC in its review of Subsection IWE, emphasis on weld-based inservice inspection was redirected toward a containment-surface inservice inspection program. Significant changes were made to accommodate this re-emphasis. The majority of these changes were published in the 1992 Edition, with the 1992 Addenda, of Subsection IWE. The NRC Proposed Rulemaking was issued for a 75-day public comment period in January, 1994. This period was extended at the request of nuclear industry organizations to allow for meaningful evaluation

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)

A Life-Cycle Risk-Informed Systems Structured Nuclear Code

International Nuclear Information System (INIS)

Hill, Ralph S. III

2002-01-01

Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The design code is a separate volume from the code for inservice inspections and both are separate from the standards for operations and maintenance. The ASME code for inservice inspections and code for nuclear plant operations and maintenance have adopted risk-informed methodologies for inservice inspection, preventive maintenance, and repair and replacement decisions. The American Institute of Steel Construction and the American Concrete Institute have incorporated risk-informed probabilistic methodologies into their design codes. It is proposed that the ASME nuclear code should undergo a planned evolution that integrates the various nuclear codes and standards and adopts a risk-informed approach across a facility life-cycle - encompassing design, construction, operation, maintenance and closure. (author)

Scheduling and coordination for in-service inspection of nuclear power plant

International Nuclear Information System (INIS)

Li Songbai

1996-11-01

Based on the practice and experiences of pre-service and in-service inspections for Daya Bay Nuclear Power Plant (NPP) by Research Institute of Nuclear Power Operation (RINPO) following RSEM code, requirements of utility and actual situation in China, the in-service inspection preparation for organization, techniques and equipment/tooling, materials, personnel and documentation is briefly described. And the scheduling and coordinating consideration for planed in-service inspection activities during NPP outage is emphatically introduced. (2 refs., 4 figs.)

Inservice Inspection in the Fugen

OpenAIRE

1983-01-01

Pressure tube type reactors have specific components orstructures,compared with light water reactors.They are(1) steam drums(2) reactor inlet headers(3) reactor inlet and outlet pipies(4) pressure tubes.Much attention is paid upon Inservice Inspection (ISI)of the above components.

Update on the EPRI power generation risk-based inservice inspection pilot plant studies

International Nuclear Information System (INIS)

Gosselin, S.R.

1997-01-01

The scope for ASME Section XI ISI programs is largely based on deterministic results contained in design stress reports. These reports are normally very conservative and may not be an accurate representation of failure potential. Service experience has shown that failures are due to either corrosion or fatigue and typically occur in areas not included in the plant's ISI program. Consequently, nuclear plants are devoting significant resources to inspection programs that provide minimum benefit. As an alternative, significant industry attention has been devoted to the application of risked-based selection criteria in order to determine the scope of inservice inspection (ISI) programs at nuclear power plants. Preliminary EPRI studies indicate that the application of these techniques will allow operating nuclear plants to reduce the examination scope of current ISI programs by as much as 60 to 80%, significantly reduce costs, and continue to maintain high nuclear plant safety standards

Development of in-service inspection plans for nuclear components at the Surry 1 nuclear power station

International Nuclear Information System (INIS)

Vo, T.V.; Simonen, F.A.; Doctor, S.R.; Smith, B.W.; Gore, B.F.

1993-01-01

As part of the nondestructive evaluation reliability program sponsored by the US Nuclear Regulatory Commission at Pacific Northwest Laboratory, a methodology has been developed for establishing in-service inspection priorities of nuclear power plant components. The method uses results of probabilistic risk assessment in conjunction with the techniques of failure modes and effects analysis to identify and prioritize the most risk-important systems and components for inspection at nuclear power plants. Surry nuclear power station unit 1 was selected for demonstrating the methodology. The specific systems selected for analysis were the reactor pressure vessel, the reactor coolant, the low pressure injection including the accumulators, and the auxiliary feedwater. The results provide a risk-based ranking of components that can be used to establish a prioritization of the components and a basis for developing improved in-service inspection plans at nuclear power plants

Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

2006-01-01

The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

International Nuclear Information System (INIS)

Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O.

2016-01-01

During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O., E-mail: vasconv@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2016-07-01

During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

From PISC to risk informed inspection

International Nuclear Information System (INIS)

Crutzen, S.; Lemaitre, P.; Fabbri, L.

1998-01-01

In this paper the correlation between inspection effectiveness, inspection qualification and risk informed inspection will be treated in view of harmonisation of inspection of nuclear plant components. Through the different phases of the PISC programme the necessity has been demonstrated to show the effectiveness of the inspection through a formal process of qualification in order to ensure that a certain level of effectiveness has been reached. Inspection qualification is indeed the way to quantify the capability level of inspection techniques used. The targets to be met by the inspection is information which should be obtained from risk based analyses. (author)

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

Multi-Canister overpack inservice inspection and maintenance

International Nuclear Information System (INIS)

SMITH, K.E.

1998-01-01

The factors to be considered in establishing inservice inspection and maintenance requirements for the Multi-Canister Overpack (MCO) include evaluating the likelihood of degradation to the MCO pressure boundary due to erosion and corrosion, reviewing commercial practice for NRC licensed spent nuclear fuel storage systems, and examining the individual MCO components for maintenance needs. Reviews of the potential for MCO erosion and corrosion conclude that neither will pose a threat to the MCO pressure boundary. Consistent with commercial practice for spent fuel storage systems, the MCO closure weld will be helium leak tested prior to placement in interim storage. Beyond the CSB facility related monitoring plans (radiological monitoring, emissions monitoring, vault cooling data, etc.), no inservice inspection or maintenance of the MCO is required during interim storage

Conceptual design of in-service inspection and maintenance for KALIMER

International Nuclear Information System (INIS)

Ju, Y. S.; Kim, S. H.; Koo, K. H.; You, B.

1999-01-01

In-service inspection and maintenance are very important for the safety and availability of nuclear power plants. The conceptual requirements of in-service inspection and maintenance should be reflected in the earlier design process for the verification of the plant operability and reliability. In this paper the fundamental approaches of the inspection and maintenance for KALIMER are established to ensure the structural integrity and operability for KALIMER. The general strategy and methodology of maintenance and inspection for the reactor system and components are proposed and described for satisfying the intents of the Section XI, Division 3, of ASME code and considering the design characteristics of KALIMER

New Swedish regulations in the area of plant inspection and in-service inspection

International Nuclear Information System (INIS)

Hansson, B.

1998-01-01

History and present status od Swedish regulations in the field of NPP inspection and in-service inspection are described. The presentation focuses on the development of regulations and establishing new ones. A description of different organisations involved is included

The use of risk indicators for establishing inspection and control priorities

International Nuclear Information System (INIS)

Alzbutas, R.; Klimasauskas, A.; Nedzinskas, L.

2003-01-01

Risk-informed approach used to support decision-making related to in-service inspections was considered in order to identify ways for establishing inspection and control priorities. The in-service inspection of piping in the Ignalina NPP has been selected as the object for research application, The inspection process studied was modelled using the integration of deterministic and probabilistic analysis methods, In order to optimize the inspection process, the focus was set on the highest risk measured as the conditional core damage frequency, It was produced by quantitatively estimated probabilities of different degradation states and conditional failure consequence probabilities, Comprehensive databases for calculation of such indicators (measures) were collected and analyzed, The failure statistical analysis as well as the evaluation of inspection efficiency expressed by the probability of defect detection was also used, The databases were further investigated with a closer attention to the data variation and comparison of risk measures using the developed software, The developed software was used to perform and administrate all the risk evaluations and ensure the possibilities to compare different options and perform sensitivity analysis, The risk measures helped to define an adequate inspection program and to focus inspections on the more important locations of the study systems, This approach allowed an optimisation of the inspection program while the probabilistic and fundamental deterministic safety requirements were maintained, The approaches to define an acceptable level of the inspection program were also considered. These approaches to define an acceptable risk were considered together with the means to reduce the number of inspection sites and the cumulative radiation exposure to the NPP inspection personnel with a reduction of overall risk, The investigated issues provided a good basis for drawing conclusions about the inspection priorities, to

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)

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.)

Status of the inservice inspection in codes; Stand der wiederkehrenden Pruefung im Regelwerk

Energy Technology Data Exchange (ETDEWEB)

Hienstorfer, W G [TUEV Energie- und Systemtechnik GmbH Baden-Wuerttemberg, Filderstadt (Germany); Hansch, M [Preussische Elektrizitaets-AG (Preussenelektra), Hannover (Germany); Koenig, G [Gemeinschaftskernkraftwerk Neckar GmbH, Neckarwestheim (Germany); Bath, H R [Kerntechnischer Ausschuss (KTA), Salzgitter (Germany)

1998-11-01

The body of regulatory nuclear engineering and inspection codes and subordinate codes defines on the one hand the general framework conditions for determination of the scope of inservice inspections, and on the other hand, for some defined components, contains detailed provisions relating to the scope of inservice tests to be made. These provisions are based on operating experience and engineering analyses. In order to establish a basis for optimization, the codes have been reviewed for possible optimization potentials, including a comparative analysis of foreign regulatory codes and approaches and the German codes. The conclusion drawn was that there is room for optimization, which should be done taking into account the following aspects: 1. Instrumentation for inservice monitoring of the state variables of relevance to the integrity of the primary loops has to be adapted, requiring acquisition of data showing local loading conditions and specific risks. 2. Inservice examinations of components and pipework should be optimized on the basis of updated, real-status information, i.e. upon availability of latest data showing verified operational loads and stresses and degradation mechanisms and risks, inspection methods, scopes and intervals can be optimized taking into account manufacturing quality. (orig./CB) [Deutsch] In dem aus gesetzlichen und untergesetzlichen Regelungen bestehenden kerntechnischen Regelwerk werden einerseits allgemeine Randbedingungen fuer die Bestimmung des Umfangs wiederkehrender Pruefungen vorgegeben und andererseits fuer einige Komponenten konkrete Festlegungen zum Pruefumfang getroffen. Diesen Festlegungen zum Pruefumfang lagen Betriebserfahrungen und ingenieurmaessige Analysen zugrunde. Als Grundlage fuer eine Optimierung der bisherigen Pruefumfaenge wurden die diesbezueglichen Moeglichkeiten des deutschen Regelwerks untersucht und ein Vergleich mit der Herangehensweise in den wichtigsten auslaendischen Regelwerken durchgefuehrt. Es zeigt

Enhanced InService Inspection system for BWR plant

International Nuclear Information System (INIS)

Shouji, Hajime; Kobayashi, Teruo; Nagao, Tetsuya

1993-01-01

An enhanced InService Inspection system (ISI) has been developed, consisting of newly designed automated scanners, automated pipe inspection system and a high speed data acquisition/analysis station. This system uses a trackless magnetic crawler type nozzle inspection scanner attached to the nozzle shoulder by magnetic wheels. The automated pipe inspection system consists of an automated scanner, controller and high speed data acquisition station. These advanced automatic inspection systems are very useful for ISI, especially in reducing the radiation exposure and time required. (author)

In-service inspections of the reactor cooling system of pressurized water reactors

International Nuclear Information System (INIS)

Fuerste, W.; Hohnerlein, G.; Werden, B.

1982-01-01

In order to guarantee constant safety of the components of the reactor cooling system, regular in-service inspections are carried out after commissioning of the nuclear power plant. This contribution is concerned with the components of the reactor cooling system, referring to the legal requirements, safety-related purposes and scope of the in-service inspections during the entire period of operation of a nuclear power plant. Reports are made with respect to type, examination intervals, examination technique, results and future development. The functional tests which are carried out within the scope of the in-service inspections are not part of this contribution. (orig.) [de

Identification of areas and time intervals for inservice inspections; Auswahl der Pruefbereiche und -intervalle fuer wiederkehrende Pruefungen

Energy Technology Data Exchange (ETDEWEB)

Metzner, K J [Preussische Elektrizitaets-AG (Preussenelektra), Hannover (Germany); Herter, K H [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Schoeckle, F [Amtec Messtechnischer Service, Lauffen (Germany)

1998-11-01

Inservice inspections are an important, redundant part of the overall concept for ensuring component integrity during NPP operation. Their efficiency however has to be considered in the context of measures creating required component quality (design and manufacture), and inspection and monitoring measures maintaining the quality of components. If an available system for inservice monitoring yields comprehensive information about real inservice loads (monitoring of the causes of degradation), inservice inspections can be designed to purposefully check the possibly resulting damaging modes and defects. If there is no or only low-level monitoring, the possible consequences of possibly damaging mechanisms that may be unknown should be checked by expanded inservice inspections plus, if necessary, unspecific, random inspections. For selection of inspection areas and determination of inspection intervals, use of design-based analyses which take into account conservative, specified loads and load frequencies (for operating modes and incidents) is not recommendable. The recommended approach is to take as a basis information on measured, real inservice loads together with information about the real quality status of components, derived from manufacturing data and service histories. The available information has to be scanned and processed so as to give a basis for new appraisal of hitherto applied NDE methods. (orig./CB) [Deutsch] Wiederkehrende Pruefungen sind ein wichtiger redundanter Teil der Gesamtmassnahmen zur Gewaehrleistung der Komponentenintegritaet waehrend des Betriebes von Kernkraftwerken. Ihre Wirksamkeit muss jedoch in enger Verbindung mit den die Komponentenqualitaet erzeugenden Massnahmen (Auslegung und Herstellung) und den die Komponentenqualitaet erhaltenden Ueberwachungmassnahmen waehrend des Betriebes gesehen werden. Wenn die vorhandene Betriebsueberwachung eine umfassende Kenntnis der bisherigen Betriebsbelastungen ermoeglicht (Ueberwachung der Ursachen

Applicability of PSA Issues for Risk Assessment during Optimisation of In-Service Inspection

International Nuclear Information System (INIS)

Kolykhanov, V.; Skalozubov, V.; Kovrigkin, Y.

2006-01-01

The current codes determining periodicity of in-service inspection of the NPP equipment have been formed using deterministic approaches and have an unnecessary degree of conservatism. A perspective direction of perfection of normative base is decision making on a basis of risk-informed methodologies. It allows to increase safety of NPP equipment's operation and to optimise programs on inspection of the equipment subject to limited resources by focusing efforts on the most safety significant elements of the equipment. It is internationally accepted that methodology of the probabilistic safety analysis (PSA) is the most universal and comprehensive tool focused on the general assessment of safety of NPP as a whole. By now, PSA Level 1 is fulfilled for all pilot units of the Ukrainian NPPs that is a valuable result, which should be taken into account at an assessment of reliability of the equipment. However, specificity of PSA methodology should be taken into account at the decision of the particular tasks aimed at optimisation of maintenance of the equipment within individual systems. The estimation of the contribution to core damage frequency (CDF) is a PSA issue usually used to assess the significance of consequences of failure of a system/equipment during risk-informed decision-making. This work shows that above factor is only a part of assessment of the significance of consequences as core damage can be expressed in different amount of the damaged fuel elements and, hence, severity of consequences. Besides CDF is directly affected only by active elements which failure can be an initiating event. PSA methodology uses averaged reliability factors of the equipment for all possible operating modes occurring at transitive accident process. Here, there are limited opportunities to account impact of periodicity of maintenance of the equipment on reliability and to predict impact of change of the inspection program. PSA methodology does not allow taking into account

French developments and experience in the field of inservice inspection

International Nuclear Information System (INIS)

Saglio, Robert; Destribats, M.-T.; Pigeon, Michel; Roule, Maurice; Touffait, A.-M.

1979-01-01

The French PWR nuclear plant program was at the origin of a large amount of R and D work in the field of inservice inspection. The actions which were undertaken may be split up into different levels: - the regulatory level, the R and D level, the design level, the flaw evaluation level. The first results of pre and inservice inspections are presented. The experience gained by French Atomic Energy Commission with new techniques like focussed ultrasonics transducers and multi frequencies Eddy current apparatus are discussed

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)

Study on in-service inspection methods for the above-ground oil tanks floors

Energy Technology Data Exchange (ETDEWEB)

Min Xiong; Yewei Kang; Mingchun, Lin; Yi Sun [PetroChina Pipeline R and D Center, Langfang (China)

2009-07-01

It is very dangerous to the environment when oil tank floors get corrosion or leak during its long-time service. The traditional inspection methods need to shut down a tank and to empty it, then to clean it in order to inspect the floor. Comparing with the traditional methods, the in-service methods can inspect tank floors rapidly without removing product and opening the tank and can save many costs of tank emptying and cleaning. This paper explores three up-to date in-service inspection methods for the oil tank floors which are acoustic emission technology ultrasonic guided wave technology and mobile robot technology. The theoretic foundation and application status of each method is described. The advantage and disadvantage of each in-service detection technology is concluded. At last some proposals are made. (author)

ASME section XI: rules for inservice inspection of nuclear power plants -an introspection

Energy Technology Data Exchange (ETDEWEB)

John, P K; Anto, Y; Mungikar, C P; Wagh, P M [Nuclear Power Corporation of India Ltd., Tarapur (India). Tarapur Atomic Power Station

1994-12-31

Section XI of the ASME BPV code is addressed to the examination, test and inspection requirements of the components of nuclear power plants (NPPs). Since its inception in 1970, this code section has undergone vast changes -probably the most among other ASME BPV code sections. Section XI is full fledged and lays down requirements with regard to all preservice inspections, inservice inspection, repair and replacement of components, tests of system etc. Tarapur Atomic Power Station (TAPS) has the distinction of being one of the earliest BWR type NPPs in the world that has an inservice inspection programme organised in line with the ASME section XI requirements. This paper summarises the experiences gained from time to time using this code section and a few suggestions to prospective users. An effort is also made to explain the philosophy of inservice inspection from ASME section XI point of view. 3 refs.

ASME section XI: rules for inservice inspection of nuclear power plants -an introspection

International Nuclear Information System (INIS)

John, P.K.; Anto, Y.; Mungikar, C.P.; Wagh, P.M.

1994-01-01

Section XI of the ASME BPV code is addressed to the examination, test and inspection requirements of the components of nuclear power plants (NPPs). Since its inception in 1970, this code section has undergone vast changes -probably the most among other ASME BPV code sections. Section XI is full fledged and lays down requirements with regard to all preservice inspections, inservice inspection, repair and replacement of components, tests of system etc. Tarapur Atomic Power Station (TAPS) has the distinction of being one of the earliest BWR type NPPs in the world that has an inservice inspection programme organised in line with the ASME section XI requirements. This paper summarises the experiences gained from time to time using this code section and a few suggestions to prospective users. An effort is also made to explain the philosophy of inservice inspection from ASME section XI point of view. 3 refs

Inservice inspection procedures and training according to the ASME code

International Nuclear Information System (INIS)

Greenwald, S.M.; Chockie, L.J.

1987-01-01

Mandatory training of the technical staff at a nuclear power plant is of paramount importance if we are to avoid costly plant shutdowns. This training should include the requirements for both Preservice and Inservice Inspection, in addition to Quality Assurance procedures as required by the American Society of Mechanical Engineers (ASME) Code. The training is best accomplished by utilizing instructors who are thoroughly familiar with plant operations and the ASME Code, as well as serving on one of the Code committees. This paper focuses on the Inservice Inspection procedures and the results of an intensive training effort to implement such procedures. (author)

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.)

In-service inspection in the Superphenix 1 vessels interspace

International Nuclear Information System (INIS)

Asty, M.; Saglio, R.

1983-03-01

The design of Superphenix 1 reactor vessels allows their in-service inspection. A self-propelling engine, the MIR, has been concieved for this need: it can do a visual and ultrasonic inspection. The MIR can move in the whole vessels interspace. The operating conditions are specified and the principle characteristics of the MIR engine are presented [fr

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

In-service inspection of nuclear power plants

International Nuclear Information System (INIS)

Asty, M.; Saglio, R.

1984-10-01

The French Commissariat a l'Energie Atomique (Atomic Energy Commission) developed two new non destructive control techniques, focused ultrasonics and multi-frequency eddy currents, which have been shown to allow a better detection and characterization of defects. We present here some of the in-service inspection devices which have been designed for field application of these techniques on the PWR reactors built by EDF, inspection devices of the PWR steam generator tubing and the now developing specific device for main tank and helicoidal tubing steam generator of Super-Phenix 1 [fr

In-service inspection of nuclear power plants

International Nuclear Information System (INIS)

1991-01-01

This Manual is intended to provide more comprehensive considerations on the management, organization, preparation, improvement and implementation of in-service inspection activities and the related surveillance. It also gives illustrative examples of good practices and recommendations from operating and other organizations that are consistent with the requirements and recommendations of the Code and Safety Guides. The Manual is directed primarily towards plant management. This Manual should be used in conjunction with the Code and the Safety Guides, in particular with IAEA Safety Series Nos. 50-C-O, 50-SG-O2, 50-SG-05, 50-SG-07, 50-SG-08 and 50-SG-D1, which contain recommendations of a general character about maintenance activities and radiation protection in an operating power plant, and with the 'Manual on the Maintenance of Systems and Components Important to Safety'. This Manual is divided into four technical sections. The first introduces the purpose, structure and main requirements of the programme. The second section describes constituents of the programme, recommending its scope, scheduling, acceptance standards and documentation of results. The following section goes into details of the inspection programme's contents, such as the selection of components, inspection locations, defect types, applicable techniques and procedures, and the evaluation of results. The last section specifies recommended methods and techniques for inspection, such as visual, ultrasonic, eddy current, magnetic particle and others. This main part of the Manual is complemented by a number of annexes which reproduce actual national examples of established procedures, ISI programme parts, acceptance standards, personnel training programmes, testing techniques and other aspects of in-service inspection, illustrating practical implementation of the recommendations of the Manual

Preventing degradation and in-service inspection of NPP steam generator

International Nuclear Information System (INIS)

Ding Xunshen

1999-01-01

The author describes the degradation of steam generator tubes in initial operating stage of France NPP. The author emphatically presents the preventive measures to tackle degradation and In-service inspection of Daya Bay NPP 1 unit as a guarantee of safety operation, including secondary side water chemistry monitoring and controlling, leakage rate monitoring, eddy-current inspection, mechanical cleaning, cleanliness inspection, foreign objects removal and etc

Laurels for Swiss in-service inspection team

International Nuclear Information System (INIS)

Anon.

1984-01-01

Following what is believed to be the first successful application of P-scan ultrasonic testing for the in-service inspection of austenitic welds in the boiling water reactor at Wuergassen (Federal Republic of Germany) last year, Sulzer participated in a 'round robin' test organized by the Electric Power Research Institute in the USA form May to July 1983. (Auth.)

A pilot application of risk-based methods to establish in-service inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station

International Nuclear Information System (INIS)

Vo, T.; Gore, B.; Simonen, F.; Doctor, S.

1994-08-01

As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish in-service inspection plans for nuclear power plant components. This method uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot applications of this method. The specific systems addressed in this report are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants. To develop inspection plans, the acceptable level of risk from structural failure for important systems and components will be apportioned as a small fraction (i.e., 5%) of the total PRA-estimated risk for core damage. This process will determine target (acceptable) risk and target failure probability values for individual components. Inspection requirements will be set at levels to assure that acceptable failure probabilistics are maintained

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

International Nuclear Information System (INIS)

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

1992-07-01

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 the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties

Examples of in-service inspections and typical maintenance schedule for low-power research reactors

International Nuclear Information System (INIS)

Boeck, H.

1997-01-01

In-service inspection methods for low-power research reactors are described which have been developed during the past 37 years of the operation of the TRIGA reactor Vienna. Special tools have been developed during this period and their application for maintenance and in-serve inspection is discussed. Two practical in-service inspections at a TRIGA reactor and at a MTR reactor are presented. Further a typical maintenance plan for a TRIGA reactor is listed in the annex. (author)

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

Aspects of the Optimization on the In-Service Inspection

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic Tuma, J.

2002-01-01

In the present paper, the aspects of optimizing In-Service Inspection (ISI) is discussed. Slovenian Nuclear Safety Administration (SNSA) and its authorized organization for the ISI activities, Institute of Metals and Technologies, are actually permanently involved in the ISI processes of the nuclear power plant (NPP) Krsko. Based on the previous experience on the ISI activities, evaluation of the results and review of the new ISI program, the decision was made to improve recent regulatory and professional practice. That means, the conclusion was made to optimize the evaluation process of the ISI as a process. Traditional criteria, standards and practice gives good fundament for the improvements implementation. Improvements can be done on the way that the more broad knowledge about safety important components of the systems shall bee added to the basic practice. It is necessary to identify conditions of the safety important components, such as realistic stress and fatigue conditions, material properties changes due aging processes, the temperature cycling effects, existing flaws characterization in the light of the previous detection and equipment technique used, assessment of the measurement accuracy on the results etc. In addition to the above mentioned, risk assessment and evaluation of the whole ISI shall be done. To do this it is necessary to made risk evaluation, based on previous structural element probability assessment. Probabilistic risk assessment is important and one of the most powerful tools in the ISI optimization. Some basic work on the filed of the risk informed methods related to the nuclear safety components has been already done. Based on reference documentation, the most important steps in risk informed ISI are discussed: scope definition, consequence evaluation, failure probability estimation, risk evaluation, non-destructive examination method selection and possibilities of implementation, monitoring and feedback. Recent experience on the ISI

Summary of inspection findings of licensee inservice testing programs at United States commercial nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Dunlop, A.; Colaccino, J.

1996-12-01

Periodic inspections of pump and valve inservice testing (IST) programs in United States commercial nuclear power plants are performed by Nuclear Regulatory Commission (NRC) Regional Inspectors to verify licensee regulatory compliance and licensee commitments. IST inspections are conducted using NRC Inspection Procedure 73756, {open_quotes}Inservice Testing of Pumps and Valves{close_quotes} (IP 73756), which was updated on July 27, 1995. A large number of IST inspections have also been conducted using Temporary Instruction 2515/114, {open_quotes}Inspection Requirements for Generic Letter 89-04, Acceptable Inservice Testing Programs{close_quotes} (TI-2515/114), which was issued January 15, 1992. A majority of U.S. commercial nuclear power plants have had an IST inspection to either IP 73756 or TI 2515/114. This paper is intended to summarize the significant and recurring findings from a number of these inspections since January of 1990.

Study and discussion on management of nuclear island in-service inspection procedure system in nuclear power plant

International Nuclear Information System (INIS)

Zhang Xueliang; Fan Yancheng

2014-01-01

In-service inspection of nuclear island is the important way for keeping safety operation of nuclear power plant. Taking Daya Bay Nuclear Power Plant as example, the management problems of in-service inspection system was studied and discussed from the angle of references, contents, classifications etc. Based on comparison with French practice, some points of view on perfection of in-service inspection system and improvement of management ability under future multi-bases and multi-units management mode were presented. (authors)

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

Risk-based inspection in the context of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Soares, Wellington A.; Vasconcelos, Vanderley de; Rabello, Emerson G., E-mail: soaresw@cdtn.br, E-mail: vasconv@cdtn.br, E-mail: egr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

Nuclear power plant owners have to consider several aspects like safety, availability, costs and radiation exposure during operation of nuclear power plants. They also need to demonstrate to regulatory bodies that risk assessment and inspection planning processes are being implemented in effective and appropriate manner. Risk-Based Inspection (RBI) is a methodology that, unlike time-based inspection, involves a quantitative assessment of both failure probability and consequence associated with each safety-related item. A correctly implemented RBI program classifies individual equipment by its risks and prioritizes inspection efforts based on this classification. While in traditional deterministic approach, the inspection frequencies are constant, in the RBI approach the inspection interval for each item depends on the risk level. Regularly, inspection intervals from RBI result in risk levels lower or equal than deterministic inspection intervals. According to the literature, RBI solutions improve integrity and reduce costs through a more effective inspection. Risk-Informed In-service Inspection (RI-ISI) is the equivalent term used in the nuclear area. Its use in nuclear power plants around world is briefly reviewed in this paper. Identification of practice methodologies for performing risk-based analyses presented in this paper can help both Brazilian nuclear power plant operator and regulatory body in evaluating the RI-ISI technique feasibility as a tool for optimizing inspections within nuclear plants. (author)

Risk-based inspection in the context of nuclear power plants

International Nuclear Information System (INIS)

Soares, Wellington A.; Vasconcelos, Vanderley de; Rabello, Emerson G.

2015-01-01

Nuclear power plant owners have to consider several aspects like safety, availability, costs and radiation exposure during operation of nuclear power plants. They also need to demonstrate to regulatory bodies that risk assessment and inspection planning processes are being implemented in effective and appropriate manner. Risk-Based Inspection (RBI) is a methodology that, unlike time-based inspection, involves a quantitative assessment of both failure probability and consequence associated with each safety-related item. A correctly implemented RBI program classifies individual equipment by its risks and prioritizes inspection efforts based on this classification. While in traditional deterministic approach, the inspection frequencies are constant, in the RBI approach the inspection interval for each item depends on the risk level. Regularly, inspection intervals from RBI result in risk levels lower or equal than deterministic inspection intervals. According to the literature, RBI solutions improve integrity and reduce costs through a more effective inspection. Risk-Informed In-service Inspection (RI-ISI) is the equivalent term used in the nuclear area. Its use in nuclear power plants around world is briefly reviewed in this paper. Identification of practice methodologies for performing risk-based analyses presented in this paper can help both Brazilian nuclear power plant operator and regulatory body in evaluating the RI-ISI technique feasibility as a tool for optimizing inspections within nuclear plants. (author)

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)

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Evaluation of effectiveness of inservice inspection in aging PWR plants based on PSA

Energy Technology Data Exchange (ETDEWEB)

Hanafusa, Hidemitsu; Irie, Takashi; Suyama, Takeshi [Institute of Nuclear Safety Systems Inc., Mihama, Fukui (Japan); Morota, Hidetsugu [Computer Software Development Co., Ltd., Tokyo (Japan)

2001-09-01

By making use of a probabilistic fracture mechanics (PFM) code which had been improved to be able to evaluate the influence of thermal aging embrittlement of cast stainless steel, we evaluated the fracture probability of welding parts of the piping of the pressurized water reactor plants in Japan and the core damage frequency by the fracture. In addition, we ranked the welding parts by the safety significance and evaluated the effectiveness of inservice inspection (ISI). As a result, it is seen that the risk increases about 25% for extending the plant life from 40 to 60 years without ISI. However, the influence of the risk on the plant life becomes negligible, when the ISI is performed adequately. This assessment is useful to plan efficient inspection and test programs of a plant under limited resources, because it is available to rank piping segments by the safety significance and to evaluate the effectiveness of inspection and test quantitatively. (author)

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.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1992-07-01

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 (LWR's); 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 the Regulatory and ASME Code 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 April 1991 through September 1991

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.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V.

1993-11-01

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 the Regulatory and ASME Code 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 programs report covering the programmatic work from April 1992 through September 1992

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

International Nuclear Information System (INIS)

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

1991-08-01

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 6 of the ASME Code. This is a progress report covering the pro grammatic work from April 1989 through September 1989. 12 refs., 4 figs. 5 tabs

An evaluation of human factors research for ultrasonic inservice inspection

International Nuclear Information System (INIS)

Pond, D.J.; Donohoo, D.T.; Harris, R.V. Jr.

1998-03-01

This work was undertaken to determine if human factors research has yielded information applicable to upgrading requirements in ASME Boiler and Pressure Vessel Code Section XI, improving methods and techniques in Section V, and/or suggesting relevant research. A preference was established for information and recommendations which have become accepted and standard practice. Manual Ultrasonic Testing/Inservice Inspection (UT/ISI) is a complex task subject to influence by dozens of variables. This review frequently revealed equivocal findings regarding effects of environmental variables as well as repeated indications that inspection performance may be more, and more reliably, influenced by the workers' social environment, including managerial practices, than by other situational variables. Also of significance are each inspector's relevant knowledge, skills, and abilities, and determination of these is seen as a necessary first step in upgrading requirements, methods, and techniques as well as in focusing research in support of such programs, While understanding the effects and mediating mechanisms of the variables impacting inspection performance is a worthwhile pursuit for researchers, initial improvements in industrial UTASI performance may be achieved by implementing practices already known to mitigate the effects of potentially adverse conditions. 52 refs., 2 tabs

An evaluation of human factors research for ultrasonic inservice inspection

Energy Technology Data Exchange (ETDEWEB)

Pond, D.J.; Donohoo, D.T.; Harris, R.V. Jr.

1998-03-01

This work was undertaken to determine if human factors research has yielded information applicable to upgrading requirements in ASME Boiler and Pressure Vessel Code Section XI, improving methods and techniques in Section V, and/or suggesting relevant research. A preference was established for information and recommendations which have become accepted and standard practice. Manual Ultrasonic Testing/Inservice Inspection (UT/ISI) is a complex task subject to influence by dozens of variables. This review frequently revealed equivocal findings regarding effects of environmental variables as well as repeated indications that inspection performance may be more, and more reliably, influenced by the workers` social environment, including managerial practices, than by other situational variables. Also of significance are each inspector`s relevant knowledge, skills, and abilities, and determination of these is seen as a necessary first step in upgrading requirements, methods, and techniques as well as in focusing research in support of such programs, While understanding the effects and mediating mechanisms of the variables impacting inspection performance is a worthwhile pursuit for researchers, initial improvements in industrial UTASI performance may be achieved by implementing practices already known to mitigate the effects of potentially adverse conditions. 52 refs., 2 tabs.

Study on in-service visual inspection using TV camera for core support graphite components in the HTTR

International Nuclear Information System (INIS)

Ishihara, Masahiro; Shibata, Taiju; Kikuchi, Takayuki; Mogi, Haruyoshi

1999-01-01

To maintain the structural integrity of graphite components during plant operation a visual inspection using a TV camera as an in-service inspection is planned in the High Temperature Engineering Test Reactor. In order to verify the in-service inspection method a preliminary analytical and experimental studies were performed. In the analytical study the harmful flaw size was determined from a viewpoint of structural integrity based on the fracture mechanics approach. Furthermore, the visible flaw size was determined by the TV camera performance test with graphite test specimens having several kinds of artificial flaws. This paper presents the analytical result on the harmful flaw size and the experimental result on the visible flaw size. From both results the applicability on the visual inspection by the TV camera as the in-service inspection is discussed in this paper. (author)

Inservice inspection of Halden BWR pressure vessel

International Nuclear Information System (INIS)

Foerli, O.; Hernes, T.

1978-01-01

A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

ASME development of risk-based inspection guidelines for nuclear power plants

International Nuclear Information System (INIS)

Gore, B.F.; Balkey, K.R.

1992-08-01

A methodology has been developed for use in preparing guidelines for the in-service inspection of nuclear power plant pressure boundary and structural components. This methodology is a further development of a general methodology previously published by the task force for application to any industry. It is unique in that it utilizes probabilistic risk assessment (PRA) information to improve the quantification of risks associated with component ruptures. A procedure has also been recommended for using the resulting quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Characteristics which an inspection strategy must possess in order to maintain target rupture probabilities are being determined by structural risk and reliability analysis (SRRA) calculations. The major features of this methodology are described in this paper, and example applications are briefly discussed. Plans for additional work in developing and applying the methodology are described

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.)

The contribution of inspection to the management of risk

Energy Technology Data Exchange (ETDEWEB)

Baron, J. [Folville Consulting Inc., Ontario (Canada)

2016-01-15

Over the recent several years, the nuclear power industry has moved to designing in-service inspection programs based on plant risk analyses. This has led to significant reduction in the amount of inspections required thereby reducing the cost of the inspection process itself but, more significantly, the cost due to plant downtime. Inspection has no impact on the consequence of failure but, rather, mitigates risk through reduction in the probability of failure. In so doing, it is important to consider the reliability of inspections, how reliability may be improved through deployment of automation but consideration must also be given to human factors with respect to achieving and reporting correct results. (author)

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

IAEA-RCA training course on in-service inspection of research reactors

International Nuclear Information System (INIS)

2002-01-01

Bhabha Atomic Research Centre (BARC) has acquired a wide range of experience for over four decades in design, construction, operation and maintenance of research reactors. The two-week training course on In-Service Inspection of Research Reactors (ISI) has greatly increased the awareness in the field of ISI of Research Reactors. The training course has been formulated so as to cover most of the topics relevant to ISI of research reactors. Important topics such as rationale for in-service inspection, material degradation mechanisms, non-destructive examination techniques, design evaluation of flaws and radiological, codal and regulatory aspects of ISI for research reactors were covered. Lectures on ISI of elastomeric materials and concrete structures, which are generally used in the construction of reactors have also been included in the course. Papers relevant to INIS are indexed separately

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

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

Risk-informed decision making a keystone in advanced safety assessment

International Nuclear Information System (INIS)

Reinhart, M.

2007-01-01

Probabilistic Safety Assessment (PSA) has provided extremely valuable complementary insight, perspective, comprehension, and balance to deterministic nuclear reactor safety assessment. This integrated approach of risk-informed management and decision making has been called Risk-Informed Decision Making (RIDM). RIDM provides enhanced safety, reliability, operational flexibility, reduced radiological exposure, and improved fiscal economy. Applications of RIDM continuously increase. Current applications are in the areas of design, construction, licensing, operations, and security. Operational phase safety applications include the following: technical specifications improvement, risk-monitors and configuration control, maintenance planning, outage planning and management, in-service inspection, inservice testing, graded quality assurance, reactor oversight and inspection, inspection finding significance determination, operational events assessment, and rulemaking. Interestingly there is a significant spectrum of approaches, methods, programs, controls, data bases, and standards. The quest of many is to assimilate the full compliment of PSA and RIDM information and to achieve a balanced international harmony. The goal is to focus the best of the best, so to speak, for the benefit of all. Accordingly, this presentation will address the principles, benefits, and applications of RIDM. It will also address some of the challenges and areas to improve. Finally it will highlight efforts by the IAEA and others to capture the international thinking, experience, successes, challenges, and lessons in RIDM. (authors)

Development of Surveillance and In-Service Inspection Programme for Indian Research Reactors Cirus and Dhruva

International Nuclear Information System (INIS)

Shukla, D.K.

2006-01-01

Many safety requirements for research reactors are quite similar to those of power reactors. For research reactors with a higher hazard potential, the use of safety codes and guides for power reactors is more appropriate. However, there are many important differences between power reactors and research reactors that must be taken into account to ensure that adequate safety margins are available in design and operation of the research reactor. Most research reactors may have small potential for hazard to the public compared to power reactors but may pose a greater potential hazard to the plant operators. The need for greater flexibility in use of research reactors for individual experiments requires a different safety approach. Safety rules for power reactors are required to be substantially modified for application to specific research reactor. Following the intent of the available safety guides for surveillance and In-Service Inspection of Nuclear Power Plants, guidelines were formulated to develop surveillance and In-Service Inspection programme for research reactors Cirus and Dhruva. Based on the specific design of these research reactors, regulatory requirements, the degree of sophistication and experience of the technical organization involved in operating the research reactor, guidelines were evolved for developing and implementing the surveillance and In-Service Inspection programme for research reactors Cirus (40 MWt) and Dhruva (100 MWt) located at Bhabha Atomic Research Centre, Trombay, Mumbai, India. Paper describes the approach adopted for formulation of surveillance and In-service Inspection programme for Dhruva reactor in detail. (author)

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

State-of-practice review of ultrasonic in-service inspection of Class I system piping in commercial nuclear power plants

International Nuclear Information System (INIS)

Morris, C.J.; Becker, F.L.

1982-08-01

The Pacific Northwest Laboratory conducted a survey to determine the state of practice of ultrasonic in-service inspection of primary system piping in light water reactors. Personnel at four utilities, five inspection organizations, and three domestic reactor manufacturers were interviewed. The intention of the study was to provide a better understanding of the actual practices employed in in-service inspection of primary system piping and of the difficulties encountered

Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

Energy Technology Data Exchange (ETDEWEB)

Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z., E-mail: raissaomarques@gmail.com, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

International Nuclear Information System (INIS)

Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z.

2017-01-01

Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

Evaluation of flaw characteristics and their influence on inservice inspection reliability

International Nuclear Information System (INIS)

Becker, F.L.

1980-01-01

This report describes the results of the first year's effort of a five year program which is being conducted by Battelle, Pacific Northwest Laboratories, on behalf of the US Nuclear Regulatory Commission. This initial effort was directed toward identification and quantification of inspection uncertainties, which are likely to occur during inservice inspection of LWR primary piping systems, and their influence on inspection reliability. These experiments were conducted on 304 stainless steel samples, however, the results are equally applicable to other materials. Later portions of the program will extend these measurements and evaluations to other materials and conditions

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)

Quality control and in-service inspection technology for hybrid-composite girder bridges.

Science.gov (United States)

2014-08-01

This report describes efforts to develop quality control tools and in-service inspection technologies for the fabrication and construction of Hybrid Composite Beams (HCBs). HCBs are a new bridge technology currently being evaluated by the Missouri De...

Methodology to identify risk-significant components for inservice inspection and testing

International Nuclear Information System (INIS)

Anderson, M.T.; Hartley, R.S.; Jones, J.L. Jr.; Kido, C.; Phillips, J.H.

1992-08-01

Periodic inspection and testing of vital system components should be performed to ensure the safe and reliable operation of Department of Energy (DOE) nuclear processing facilities. Probabilistic techniques may be used to help identify and rank components by their relative risk. A risk-based ranking would allow varied DOE sites to implement inspection and testing programs in an effective and cost-efficient manner. This report describes a methodology that can be used to rank components, while addressing multiple risk issues

Improved in-service inspection program for management of degradation in steam generator tubing

International Nuclear Information System (INIS)

Kurtz, R.; Heasler, P.; Muscara, J.

1992-01-01

This paper presents an overview of significant results from NRC-sponsored research on steam generator tube integrity and inspection. Burst test results are described along with empirical models to relate flaw geometry and size to tube burst pressure. Results of round robin examinations of a retired-from-service steam generator to determine eddy current inspection reliability are presented. An evaluation and comparison of various sampling plans for in-service inspection of steam generators is discussed. Finally, performance demonstration qualification efforts for eddy current inspection systems are described

Conceptual design and assessment of in-service inspection and maintenance of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Seok Hun; Kim, Jong Bum; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-05-01

In the conceptual design stage of KALIMER, the philosophy and methodology of in-service inspection (ISI) and maintenance for the reactor system and components are proposed and described. The ISI and maintenance should be carried out throughout plant life to ensure the structural integrity and safety of KALIMER. The conceptual design of ISI and maintenance are performed for considering the design characteristics of KALIMER and the intents of the ASME XI Division 3. This report describes and summarizes the requirements and available methods of ISI and maintenance. The visual inspection and continuous monitoring play a great role in the in-service inspection of KALIMER. The major structures of KALIMER reactor system are designed for maintenance free operation for the plant life time and the maintenance philosophy is to replace major components rather than repair them. The assessment of the ISI accessibility and maintainability is performed and reviewed each major component. The postulated failure defects for each component are estimated and evaluated for KALIMER safety and reliability. 8 refs., 16 figs., 13 tabs. (Author)

Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

Energy Technology Data Exchange (ETDEWEB)

Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2003-03-01

Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

Evaluation and improvement in 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.

1988-01-01

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 NRC 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 October 1986 through September 1987. (author)

Evaluation and improvement in 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.

1988-01-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactor (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC 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 October 1986 through September 1987

77 FR 69508 - Inservice Inspection of Prestressed Concrete Containment Structures With Grouted Tendons

Science.gov (United States)

2012-11-19

... Containment Structures With Grouted Tendons AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide... (RG) 1.90, ``Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons... appropriate surveillance program for prestressed concrete containment structures with grouted tendons...

GIMIS - Integral Solution for the In-Service Inspection Management of Components in NPPs

International Nuclear Information System (INIS)

Borovic, Damir; Vukovic, Igor

2014-01-01

Performance of in-service testing and inspection of components and systems in nuclear power plants are required in order to maintain the nuclear power plant while in operation and to return the plant to service, following plant outages. GIMIS is comprehensive software that integrates all processes, functions and data related to planning, administrating and executing inspections on systems, structures and components in nuclear power plants. The software is designed as a web application developed using the Microsoft ASP.NET technology, database is a Microsoft SQL, and client is composed of JavaScript frameworks. It can be adapted to local language, regulations, and requirements according to the power plant needs. The application consists of seven interdependent modules, namely: components, equipment, personnel, requirements, inspection planning, inspection execution, and reports. GIMIS deals with component, equipment and personnel management (both internally employed and outsourced), also provides full component history including uploaded documentation, drawings, previous inspection results, and supports various types of requirements. It enables scheduling facilitation using component data and compliance requirements, offers generation of all inspection relevant documentation and reports, and covers the equipment management including calibration requirements, certification of equipment and allocation to specific inspections/outages, as well as personnel certifications and allocation to specific inspections/outages. The paper describes the content and functionality of the GIMIS application and provides information of its built-in capabilities and features. (authors)

Review of In-Service Inspection and Repair Technique Developments for French Liquid Metal Fast Reactors

International Nuclear Information System (INIS)

Baque, F.

2005-01-01

In-service monitoring of nuclear plants is indispensable for both the Operator and the Regulator. The notion of in-service monitoring ranges from the continuous monitoring of the reactor in operation to the thorough in-service reactor inspection during programmed shutdowns. However, the highly specific environment found in French liquid metal fast reactor plants - Phenix and Superphenix - makes monitoring and inspection complicated because of the use of a sodium coolant that is hot, opaque, and difficult to drain.The Commissariat a l'Energie Atomique, in collaboration with its traditional French partners, Electricite de France utilities and FRAMATOME/Novatome Engineering, decided to conduct a 6-yr research and development program (1994-2000) to explore this problem vis-a-vis Superphenix, as well as the possibilities of intervening within the reactor block or on components in a sodium environment. Furthermore, the safety reevaluation of Phenix, conducted between 1994 and 2003, represented an excellent 'test bench' during which the limits of inspection processes - applied to an integrated reactor concept - were surpassed using techniques such as fuel subassembly head scanning, ultrasonic examination of the core support, and visual inspection of the cover-gas plenum following a partial sodium draining. Repair techniques were investigated for cleaning of sodium wet structure surfaces, cutting of damaged parts, and welding in sodium aerosol atmosphere. Both conventional and laser processes were tested

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

Experience with the implementation of a risk-based ISI program and inspection qualification

International Nuclear Information System (INIS)

Chapman, O.J.V.

1996-01-01

Rolls Royce and Associates (RRA) are the Design Authority (DA) for Nuclear Steam Raising Plant (NSRP) used for the Royal Naval Nuclear Fleet. Over the past seven years RRA, with support from the Ministry of Defense, has developed and implemented a risk based in-service inspection (RBISI) strategy for the NSRP. Having used risk as a means of optimizing where to inspect, an inspection qualification (IQ) process has now been put in place to ensure that proposed inspections deliver the expected gains assumed. This qualification process follows very closely that currently being put forward by the European Network on Inspection Qualification (ENIQ)

Experience with ANSI N14.30 for in-service inspections of semi-trailer used for spent fuel shipments

International Nuclear Information System (INIS)

Hirtz, G.J.

1998-05-01

On July 18, 1996, the Oak Ridge National Laboratory (ORNL) resumed shipping spent fuel in interstate commerce after a 10 year suspension of this activity. This shipment was conducted using a Nuclear Regulatory Commission licensed spent fuel transport package purchased from General Electric Company by ORNL for the purpose of moving High Flux Isotope Reactor spent fuel to the Savannah River Site. The trailer, fabricated to the ANSI N14.30, Semi-Trailers Employed in the Highway Transport of Weight Concentrated Radioactive Loads Design, Fabrication, and Maintenance, has recently undergone its first scheduled in-service inspection. This paper presents the experience gained from interpretation and application of the ANSI N14.30 standard focusing on the in-service inspection for the structure of the trailer. Initially, the term weight concentrated is illustrated giving detail to the location and center of gravity of the 33,500 pound shipping container and forces induced by the tie down system. Basic information about the design stresses and initial testing provided by the manufacturer are used as a lead-in to the requirements of the standard. The task of examining the trailer structure provided many lessons and required considerable effort. All of the support personnel were provided by ORNL; the garage mechanics and the certified inspection engineers had never been involved in applying ANSI N14.30. Other obstacles were the lack of existing inspection procedures for this particular activity and the lack of a previous experience interpreting the standard with regard to repair work. Some of these questions were resolved by clarification received from the writers of the standard, and others were resolved by the teamwork between the manufacturer and ORNL. This experience illustrated the importance of the trailer manufacturer as a participant in the decisions made concerning in-service inspection and maintenance

Development of data acquisition and processing system for In-service inspection of nuclear power plants

International Nuclear Information System (INIS)

Takama, Shinkichi; Kobayashi, Koji; Satoh, Yoshio; Koga, Yoshihiro; Shimizu, Takakazu

1981-01-01

In-service inspection (ISI) is required during the plant outage to assure the reliability of the components of a nuclear power plant. IHI has developed the advanced ISI system which consists of remote controlled and mechanized ultrasonic inspection devices for reactor pressure vessel examination, semi-automatic inspection instruments for piping examination with manual scan and automatic recording, data acquisition and processing system with microprocessor and mini-computer. By this system, ISI can be performed fully satisfying the requirement of ASME Code Sec. XI and minimizing operation in the high radioactive areas. All ultrasonic information is processed by the computer and the examination results such as size and location of ultrasonic indication are printed out in the form of sectional and plan view of the part examined, reproduced screen image and polar plot, etc. as well as the evaluation sheet. This system saves the elaborate work of inspection personnel and is expected to contribute to the improvement of inspection quality and to the reduction of radiation exposure of inspection personnel. (author)

Specialists' meeting on in-service inspection and monitoring of LMFBRs. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-11-01

The purpose of the meeting sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors (IWGFR) was attended by participants from France, the Federal Republic of Germany, Italy, Japan, the Netherlands, United Kingdom, United States of America, and two international organizations: CEC and IAEA. The purpose of the meeting was to review and discuss inspection requirements and concepts developed in the IWGFR member countries, development of inspection methods and their application in practical work and to recommend future development. The technical sessions were divided into four topical sessions as follows: General Concepts for In-Service Inspection of the Reactor Plant; Remotely Controlled Inspection Systems; Ultrasonic Test Methods; Miscellaneous Test Methods.

Specialists' meeting on in-service inspection and monitoring of LMFBRs. Summary report

International Nuclear Information System (INIS)

1980-11-01

The purpose of the meeting sponsored by the IAEA on the recommendation of the International Working Group on Fast Reactors (IWGFR) was attended by participants from France, the Federal Republic of Germany, Italy, Japan, the Netherlands, United Kingdom, United States of America, and two international organizations: CEC and IAEA. The purpose of the meeting was to review and discuss inspection requirements and concepts developed in the IWGFR member countries, development of inspection methods and their application in practical work and to recommend future development. The technical sessions were divided into four topical sessions as follows: General Concepts for In-Service Inspection of the Reactor Plant; Remotely Controlled Inspection Systems; Ultrasonic Test Methods; Miscellaneous Test Methods

Nuclear containment systems and in-service inspection status of Korea nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Jihong, Park; Jaekeun, Hong; Banuk, Park [Korea Institute of Machinery and Materials, Dept. of Authorized Test and Evaluation, Kyungnam (Korea, Republic of)

2007-07-01

20 unit nuclear power plants in Korea have been operated and maintained since the first unit started in commercial service in 1978. Most recently 4 units were under construction and several units were planned to be constructed. by industries. 4 types of nuclear containment systems have been constructed until now: first, metal containments, then pre-stressed concrete containments with grouted tendon systems, followed by pre-stressed concrete containments with un-grouted tendon systems, and Korea standard nuclear containments. All the nuclear containments should be inspected periodically. Therefore for periodic in-service inspection, several appropriate technical requirements should be applied differently depending on the specific nuclear containment types. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally nuclear containment system of Korea standard nuclear power plant was settled down, and as a matter of course it dominates the trend of present and future nuclear containment systems. Overall in-service inspection results of most Korea nuclear containments have not showed any serious evidence of degradation.

Review of Savannah River Site K Reactor inservice inspection and testing restart program

International Nuclear Information System (INIS)

Anderson, M.T.; Hartley, R.S.; Kido, C.

1992-09-01

Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

Risk-informed inservice test activities at the NRC

International Nuclear Information System (INIS)

Fischer, D.; Cheok, M.; Hsia, A.

1996-01-01

The operational readiness of certain safety-related components is vital to the safe operation of nuclear power plants. Inservice testing (IST) is one of the mechanisms used by licensees to ensure this readiness. In the past, the type and frequency of IST have been based on the collective best judgment of the NRC and industry in an ASME Code consensus process and NRC rulemaking process. Furthermore, IST requirements have not explicitly considered unique component and system designs and contribution to overall plant risk. Because of the general nature of ASME Code test requirements and non-reliance on risk estimates, current IST requirements may not adequately emphasize testing those components that are most important to safety and may overly emphasize testing of less safety significant components. Nuclear power plant licensees are currently interested in optimizing testing by applying resources in more safety significant areas and, where appropriate, reducing measures in less safety-significant areas. They are interested in maintaining system availability and reducing overall maintenance costs in ways that do not adversely affect safety. The NRC has been interested in using probabilistic, as an adjunct to deterministic, techniques to help define the scope, type and frequency of IST. The development of risk-informed IST programs has the potential to optimize the use of NRC and industry resources without adverse affect on safety

Risk-informed inservice test activities at the NRC

Energy Technology Data Exchange (ETDEWEB)

Fischer, D.; Cheok, M.; Hsia, A.

1996-12-01

The operational readiness of certain safety-related components is vital to the safe operation of nuclear power plants. Inservice testing (IST) is one of the mechanisms used by licensees to ensure this readiness. In the past, the type and frequency of IST have been based on the collective best judgment of the NRC and industry in an ASME Code consensus process and NRC rulemaking process. Furthermore, IST requirements have not explicitly considered unique component and system designs and contribution to overall plant risk. Because of the general nature of ASME Code test requirements and non-reliance on risk estimates, current IST requirements may not adequately emphasize testing those components that are most important to safety and may overly emphasize testing of less safety significant components. Nuclear power plant licensees are currently interested in optimizing testing by applying resources in more safety significant areas and, where appropriate, reducing measures in less safety-significant areas. They are interested in maintaining system availability and reducing overall maintenance costs in ways that do not adversely affect safety. The NRC has been interested in using probabilistic, as an adjunct to deterministic, techniques to help define the scope, type and frequency of IST. The development of risk-informed IST programs has the potential to optimize the use of NRC and industry resources without adverse affect on safety.

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.)

On-line monitoring and inservice inspection in codes

International Nuclear Information System (INIS)

Bartonicek, J.; Zaiss, W.; Bath, H.R.

1999-01-01

The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

Applications of risk informed principles at BKAB

International Nuclear Information System (INIS)

Ingemarson, I.

1999-01-01

BKAB has developed a detailed and useful model for PSA Level 1 that was finished in the beginning of 1999. The PSA-model has already been used in several activities as a tool to form the basis for risk-informed decisions. First of all it is used to reduce the risk that have been identified in the PSA Level 1 but there are plants to use it in justification of criteria in the technical specification, in in-service inspection with focus on the reactor pressure boundary piping and in classifying critical components. (au)

In-service inspection of fertilizer and steel and ship building industries

International Nuclear Information System (INIS)

Sanaullah, Md.

1998-01-01

There is a good number of fertilizer factories, and steel and ship building industries in Bangladesh. Different NDT methods are being used for in-service inspection of these industries to ensure product reliability, to prevent accident and save human life, to maintain a uniform quality level, to lower manufacturing cost and to minimize the wastage of time and materials and for other benefits. (author)

The Mechatronic System Design Of Ultrasonic Scanner For Inservice Inspection Of Research Reactor

Science.gov (United States)

Handono, Khairul; Kristedjo, K.; Awwaluddin, M.; Shobary, Ihsan

2018-02-01

The mechatronic system design of ultrasonic scanner for inservices inspection of Research Reactor has been conducted. The requirement designed must be reliable operated, safety to personnel and equipments, ease of maintenance and operation, protection of equipment mechanically, interchangeability of equipments and addition of the several model of probe immersion ultrasonic tranducer. In order to achieve the above goals and obtain the desired results, a mechatronic design based on mechanical and electronic practical experiences will be needed. In this paper consist of the mechanical design and the system mechanical movement using stepper motor control. The criteria and the methods of designs of mechanical and electronic equipments of the system have been discussed and investigated. A mechanical and instrumentation control system drawing and requirement of design will be presented as the outcome of the design. The designed of mechanical system is consequently simulated by solidwork software. The intention of the above research is to create solutions in different ways of inservice inspection of integrity of Reactor.

New approach for risk based inspection of H2S based Process Plants

International Nuclear Information System (INIS)

Vinod, Gopika; Sharma, Pavan K.; Santosh, T.V.; Hari Prasad, M.; Vaze, K.K.

2014-01-01

Highlights: • Study looks into improving the consequence evaluation in risk based inspection. • Ways to revise the quantity factors used in qualitative approach. • New approach based on computational fluid dynamics along with probit mathematics. • Demonstrated this methodology along with a suitable case study for the said issue. - Abstract: Recent trend in risk informed and risk based approaches in life management issues have certainly put the focus on developing estimation methods for real risk. Idea of employing risk as an optimising measure for in-service inspection, termed as risk based inspection, was accepted in principle from late 80s. While applying risk based inspection, consequence of failure from each component needs to be assessed. Consequence evaluation in a Process Plant is a crucial task. It may be noted that, in general, the number of components to be considered for life management is very large and hence the consequence evaluation resulting from their failures (individually) is a laborious task. Screening of critical components is usually carried out using simplified qualitative approach, which primarily uses influence factors for categorisation. This necessitates logical formulation of influence factors and their ranges with a suitable technical basis for acceptance from regulators. This paper describes application of risk based inspection for H 2 S based Process Plant along with the approach devised for handling the influence factor related to the quantity of H 2 S released

Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

International Nuclear Information System (INIS)

Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

1990-01-01

Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective of this paper is to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems

Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

International Nuclear Information System (INIS)

Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

1990-03-01

Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective was to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems. 2 refs., 4 figs., 5 tabs

Maintenance, surveillance and in-service inspection in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

Effective maintenance, surveillance and in-service inspection (MS and I) are essential for the safe operation of a nuclear power plant. The objective of this Safety Guide is to provide recommendations and guidance for MS and I activities to ensure that SSCs important to safety are available to perform their functions in accordance with the assumptions and intent of the design. This Safety Guide covers the organizational and procedural aspects of MS and I. However, it does not give detailed technical advice in relation to particular items of plant equipment, nor does it cover inspections made for and/or by the regulatory body. This Safety Guide provides recommendations and guidance for preventive and remedial measures, including testing, surveillance and in-service inspection, that are necessary to ensure that all plant structures, systems and components (SSCs) important to safety are capable of performing as intended. This Safety Guide covers measures for fulfilling the organizational and administrative requirements for: establishing and implementing schedules for preventive and predictive maintenance, repairing defective plant items, selecting and training personnel, providing related facilities and equipment, procuring stores and spare parts, and generating, collecting and retaining maintenance records for establishing and implementing an adequate feedback system for information on maintenance. MS and I should be subject to quality assurance in relation to all aspects important to safety. Quality assurance has been dealt with in detail in other IAEA safety standards and is covered here only in specific instances, for emphasis. In Section 2, a concept of MS and I is presented and the interrelationship between maintenance, surveillance and inspection is discussed. Section 3 concerns the functions and responsibilities of different organizations involved in MS and I activities. Section 4 provides recommendations and guidance on such organizational aspects as

Maintenance, surveillance and in-service inspection in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

Effective maintenance, surveillance and in-service inspection (MS and I) are essential for the safe operation of a nuclear power plant. The objective of this Safety Guide is to provide recommendations and guidance for MS and I activities to ensure that SSCs important to safety are available to perform their functions in accordance with the assumptions and intent of the design. This Safety Guide covers the organizational and procedural aspects of MS and I. However, it does not give detailed technical advice in relation to particular items of plant equipment, nor does it cover inspections made for and/or by the regulatory body. This Safety Guide provides recommendations and guidance for preventive and remedial measures, including testing, surveillance and in-service inspection, that are necessary to ensure that all plant structures, systems and components (SSCs) important to safety are capable of performing as intended. This Safety Guide covers measures for fulfilling the organizational and administrative requirements for: establishing and implementing schedules for preventive and predictive maintenance, repairing defective plant items, selecting and training personnel, providing related facilities and equipment, procuring stores and spare parts, and generating, collecting and retaining maintenance records for establishing and implementing an adequate feedback system for information on maintenance. MS and I should be subject to quality assurance in relation to all aspects important to safety. Quality assurance has been dealt with in detail in other IAEA safety standards and is covered here only in specific instances, for emphasis. In Section 2, a concept of MS and I is presented and the interrelationship between maintenance, surveillance and inspection is discussed. Section 3 concerns the functions and responsibilities of different organizations involved in MS and I activities. Section 4 provides recommendations and guidance on such organizational aspects as

An automated eddy current in-service inspection system for nuclear steam generator tubing

International Nuclear Information System (INIS)

Wells, N.S.

1981-06-01

A prototype steam generator in-service inspection system incorporating remotely-controlled instrumentation linked by a digital transmission line to an instrument and control trailer outside the reactor containment has been designed and manufactured and is presently undergoing field tests. The (Monel 400) steam generator tubes are scanned two at a time using absolute eddy current probes controlled by two remotely-operated probe drives at a scanning speed of 0.5 m/s. The probes are positioned on the tubesheet by a light-weight (1.5 kg) microprocessor-operated tubesheet walker mechanism. Digitized control and data signals are transmitted up to 300 m to the control trailer. There the control and analysis computers extract the relevant signal information and present it in condensed form as labelled graphics on CRT consoles for on-line visual assessment. Hard copy output is also provided for each tube scanned (one per minute). Condensed data is archived on magnetic tapes for additional off-line analysis and comparisons with other inspections

An expert panel approach to support risk-informed decision making

International Nuclear Information System (INIS)

Pulkkinen, U.; Simola, K.

2000-01-01

The report describes the expert panel methodology developed for supporting risk-informed decision making. The aim of an expert panel is to achieve a balanced utilisation of information and expertise from several disciplines in decision-making including probabilistic safety assessment as one decision criterion. We also summarise the application of the methodology in the STUK's RI-ISI (Risk-Informed In-Service Inspection) pilot study, where the expert panel approach was used to combine the deterministic information on degradation mechanisms and probabilistic information on pipe break consequences. The expert panel served both as a critical review of the preliminary results and as a decision support for the final definition of risk categories of piping. (orig.)

Impact of inservice inspection on the reliability of nuclear piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-12-01

The reliability of nuclear piping is a function of piping quality as fabricated, service loadings and environments, plus programs of continuing inspection during operation. This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident

Facilities for in-service control

International Nuclear Information System (INIS)

Werner, H.

1980-01-01

Up to now the efficiency of in-service control had been limited and dependent on special qualifications of the super-intending personnel and was thus neither exact nor transferable. In the meantime a great number of testing and measuring facilities for special application have been developed. Even novel types of testing methods do not grant absolutely precise statements. In most cases, however, there is a possibility of planning repairs by trend control and of avoiding failures of operation with aggravating consequences. Technical designers from all subject fields should be made familiar with the continually increasing spectrum of in-service inspection techniques so that a better application of modern and well-tried testing methods can be planned. Systematic in-service inspections complete the Rules of Technology which are safety-oriented with regard to the nonlinear wear between regular dates of control which has been ignored so far and its implied risks. Components which are not liable to control can be checked much better by novel type methods. (orig./RW) [de

Risk-based optimization of pipe inspections in large underground networks with imprecise information

International Nuclear Information System (INIS)

Mancuso, A.; Compare, M.; Salo, A.; Zio, E.; Laakso, T.

2016-01-01

In this paper, we present a novel risk-based methodology for optimizing the inspections of large underground infrastructure networks in the presence of incomplete information about the network features and parameters. The methodology employs Multi Attribute Value Theory to assess the risk of each pipe in the network, whereafter the optimal inspection campaign is built with Portfolio Decision Analysis (PDA). Specifically, Robust Portfolio Modeling (RPM) is employed to identify Pareto-optimal portfolios of pipe inspections. The proposed methodology is illustrated by reporting a real case study on the large-scale maintenance optimization of the sewerage network in Espoo, Finland. - Highlights: • Risk-based approach to optimize pipe inspections on large underground networks. • Reasonable computational effort to select efficient inspection portfolios. • Possibility to accommodate imprecise expert information. • Feasibility of the approach shown by Espoo water system case study.

On the approximation of crack shapes found during inservice inspection

International Nuclear Information System (INIS)

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

1997-01-01

This paper addresses the characterization of axial internal flaw found during inservice inspection of a pipe. J-integral distribution for various flaw shapes is obtained using line spring finite, element method. The peak J-value and its distribution across the crack is found to be characteristic feature of each shape. The triangular shape yields peak J-value away from the center, the point of depth. The elliptic approximation results in large overestimate of J-value for unsymmetric flaws. Triangular approximation is recommended for such flaws so that further service can be obtained from the component

On the approximation of crack shapes found during inservice inspection

Energy Technology Data Exchange (ETDEWEB)

Bhate, S.R.; Chawla, D.S.; Kushwaha, H.S. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01

This paper addresses the characterization of axial internal flaw found during inservice inspection of a pipe. J-integral distribution for various flaw shapes is obtained using line spring finite, element method. The peak J-value and its distribution across the crack is found to be characteristic feature of each shape. The triangular shape yields peak J-value away from the center, the point of depth. The elliptic approximation results in large overestimate of J-value for unsymmetric flaws. Triangular approximation is recommended for such flaws so that further service can be obtained from the component.

Review of international developments and cooperation on Risk-Informed In-Service-Inspection (RI-ISI) and Non-destructive Testing (NDT) Qualification in OECD-NEA member countries- Responses to the questionnaire - CSNI/integrity and ageing working group

International Nuclear Information System (INIS)

2005-01-01

In December 2000, the Committee on Nuclear Regulatory Activities (CNRA) and the Committee on the Safety of Nuclear Installations (CSNI) agreed to prepare a state-of-the art report addressing the present situation and regulatory aspects in NEA member countries on: - Risk based / risk informed in-service inspections (ISI) developments, - Qualification of NDT system to be used for the inspections. The CSNI gave mandate to the CSNI working group on the Integrity of Components and Structures (IAGE) to prepare the report. In order to get a good basis for compiling the report with an overview on the present situation in OECD countries and regulatory aspects on the further developments of RI-ISI and NDT qualification approaches a questionnaire was prepared. This questionnaire was organised in two parts. The first part addressed used risk based / risk informed ISI approaches and regulatory aspects on the further developments. The second part addressed used NDT qualification approaches and other measures for getting reliable inspection results as well as regulatory aspects on the further developments of qualification approaches. Some parts of the questionnaire addressed topics, which have been dealt with in other European or national programs. Available relevant information from these programs has been also collected. The questionnaire was circulated in 2003 among NEA member countries organisations. Appendix 1 contains the questionnaire. Appendix 2 contains the compilation of responses to the questionnaire. A workshop was organized to complement the questionnaire (NEA/CSNI/R(2004)9 Proceedings of the CSNI Workshop on 'International developments and cooperation on Risk-Informed In-Service- Inspection (RI-ISI) and Non-destructive Testing (NDT) Qualification' held in Stockholm, Sweden on 13-14 April 2004 and hosted by SKI). In addition to regulators, licensees, manufacturers and researchers, this workshop gathered international organisations (i.e. EC, IAEA) and the main

The Information-Seeking Habits of In-Service Educators

Science.gov (United States)

Shipman, Todd; Bannon, Susan H.; Nunes-Bufford, Kimberly

2015-01-01

Research on information literacy and educators has focused on preservice educators and learning information literacy skills. Little research exists on in-service educators and their information literacy skills. Purposes of this study were to identify information sources that in-service educators used; to determine relationships between information…

Improvement of in-service inspection in nuclear power plants

International Nuclear Information System (INIS)

2004-07-01

During the nuclear power plant's (NPPs) operating lifetime, the operating organisation should examine the systems, structures and components (SSCs) for possible deterioration so as to determine whether they are acceptable for continued safe operation or whether remedial measures should be taken. In-service inspection (ISI) provides the systematic framework for these examinations. An effective ISI programme ensures both that the safety of the plant is not adversely affected after the commencement of operation and that the levels of reliability and availability of all plant SSCs remain in accordance with the assumptions and intent of the design and, consequently, a cost effective electricity generation is guaranteed. This report was prepared under one of the IAEA projects, 'Integrated NPP life cycle management including decommissioning'. The main objective of an integrated life cycle management programme is to enable NPPs to compete, without compromising safety, successfully in the changing energy markets throughout their service life and to facilitate life extension and eventual decommissioning through improved engineering, technological, economic and managerial actions. The Technical Working Group on NPP Life Management and other advisory groups nominated by the Member States provide recommendations on high priority needs of Member States in this area. The purpose of this TECDOC is to: discuss and evaluate the status of ISI and its evolution in NPPs of IAEA Member States; discuss and evaluate the criteria for effective ISI and the constitutive elements; discuss and evaluate the implication of the introduction of risk informed ISI (RI-ISI) concepts; generate a common recommendation as to how to benefit from utilising the potential of RI-ISI concepts for further development and, possibly, improvement of ISI effectiveness. The specific concept of this report is to treat the two key issues, i.e. inspection qualification and RI-ISI, and their complementarities, to

Current products and future plan of regulatory research for risk-informed regulation in Korea

International Nuclear Information System (INIS)

Sung, Key Yong; Lee, Chang Ju; Kim, Woong Sik; Kim, Hho Jung

2003-01-01

The first phase of a regulatory research project for risk-informed regulation (RIR) and applications (RIA) was finished in March of 2002. Various results that could be useful for preparing Korean RIR system have been developed. One of the remarkable outputs is development of reactor safety goals and acceptance criteria for RIR and RIA in Korea. The Safety Goal has a 4-tier hierarchical structure and each tier has specified goals classified for their usage. Regulatory review guides for probabilistic safety assessment (PSA) including level-1, level-2 and low power and shutdown PSA have been updated by reflecting new information obtained from not only the overseas documents but also experience and insights from regulatory review in Korea. In addition, draft regulatory guides for risk-informed in-service inspection, in-service testing, importance ranking of motor-operated valves, and AOT/STI change of Technical Specifications have been developed for preparing ongoing and future licensing work. Risk-based inspection guides with inspection items selected from a viewpoint of risk importance have been suggested for Korean standard NPPs as well. In the second phase of a research project (April of 2002 to March of 2005), two regulatory research projects on RIR were initiated. One is a study on institutionalization of risk-informed and performance-based regulation. Main topics of this project are evaluation of benefit and characteristics of RIR, development of optimized Korean RIR model, impact analysis for the change of current regulation framework, and suggestion of RIR-related laws and rules. The other is focusing on the development in the areas of a regulatory audit PSA model and regulatory guides for risk monitoring, and application techniques of risk information to the significance determination of plant performance indicators and inspection findings. It is expected that a concrete scheme and detailed regulatory techniques for embodiment of RIR system in Korea will be

Beliefs concerning the reliability of nuclear power plant in-service inspections

International Nuclear Information System (INIS)

Kettunen, J.

1997-01-01

The aim of the study was to investigate belief systems held by the officials responsible for the planning and supervision of NDT operations within the Finnish nuclear industry. They were asked to express their opinions on (1) the reliability of NDT methods in general, (2) the factors influencing the reliability of in-service inspections, and (3) the degree of reliability of the current inspections operations conducted by means of NDT methods in the Finnish nuclear power plants. Another goal of the study was to assess the adequacy of officials' beliefs (or belief systems). The research data was collected by interviewing representatives from Finnish power companies (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations, and the Finnish Centre for Radiation and Nuclear Safety (STUK). The adequacy of the beliefs expressed was assessed by means of the results obtained from international NDT reliability studies and on the basis of interviewees' own justification. (refs.)

Proposals for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors

International Nuclear Information System (INIS)

Bolt, P.R.

1976-01-01

Design and operational experience of CEGB gas cooled reactors and certain overseas reactor plant is reviewed in relation to in-service inspection and monitoring capabilities. Design guidelines and preliminary proposals are given for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors. Specific comments are made on the items of further design and development work believed to be necessary

Human reliability impact on in-service inspection

International Nuclear Information System (INIS)

Spanner, J.C. Sr.

1986-01-01

This paper describes a study conducted to identify, characterize, and evaluate the human reliability aspects of ultrasonic testing/inservice inspection (UT/ISI). Recent measurements of UT/ISI system effectiveness have revealed wide variations in performance; suggesting that insufficient emphasis is being placed on the human reliability aspects of nondestructive examination. It appears that NDE performance can be improved through application of the human factors principles relating to the task, training, procedure, environmental, and individual difference variables. These variables are collectively referred to as performance-shaping factors. A man-machine systems model was developed to describe the UT/ISI process using functional task descriptors. The relative operating characteristic (ROC) analysis method, which is derived from signal detection theory, offers unique attributes for analyzing NDT performance. The results of a limited human factors evaluation conducted in conjunction with a mini-round robin test are also described

Automatic ultrasonic testing and the LOFT in-service inspection program

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

An automatic ultrasonic testing system has been developed which significantly improves the flaw indication detection and characterization capability over the capability of conventional volumetric examination techniques. The system utilizes an accurately located ultrasonic sensor to generate the examination data. A small computer performs and integrates control and data input/output functions. Computer software has been developed to provide a rigorous method for data analysis and ultrasonic image interpretation. The system has been used as part of an in-service inspection program to examine welds in thich austenitic stainless steel pipes in a small experimental nuclear reactor

Application of risk-based methods for inspection of nuclear power plant components

International Nuclear Information System (INIS)

Balkey, K.R.

1992-01-01

In-service inspections (ISIs) can play a significant role in minimizing equipment and structural failures. All aspects of inspections, i.e., objectives, method, timing, and the acceptance criteria for detected flaws can affect the probability of component failure. Where ISI programs exist, they are primarily based on prior experience and engineering judgment. At best, some include an implicit consideration of risk (probability of failure multiplied by consequence). Since late 1988, a multidisciplined American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection Guidelines has been addressing the general question of how to formally incorporate risk considerations into plans and requirements for the ISI of components and structural systems. The task force and steering committee that guided the project have concluded that appropriate analytical methods exist for evaluating and quantifying risks associated with pressure boundary and structural failures. With the support of about a dozen industry and government organizations, the research group has recommended a general methodology for establishing a risk-based inspection program that could be applied to any nuclear system or structural 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

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

Report on the Regulators Experience of NDT Qualification for In-service Inspection of Nuclear Components

International Nuclear Information System (INIS)

2003-08-01

several times, and first in 1999 the main part of the pilot study was completed. However, results from one important part of the pilot study are still missing. This relates to a planned comparison with results from an in-service inspection simulation on cracked piping parts that had been removed from a nuclear power plant. The objective of this third task was to compile a status report with information of actions taken, national developments, and experience since the publication of the common position document EUR 16802 in 1997, to follow the second ENIQ Pilot Study and evaluate the results from a regulatory point of view. The experience exchange started with a questionnaire that was sent to the present task force members 6 and to BCCN in France, who decided to not actively participate in the third task of the Task Force. The present report represents the work product of the experience exchange activities conducted by the Task Force. (author)

Development and validation of real-time SAFT-UT system for inservice inspection of LWRs

International Nuclear Information System (INIS)

Doctor, S.R.; Hall, T.E.; Reid, L.D.; Mart, G.A.

1988-01-01

The Pacific Northwest Laboratory is working to design, fabricate, and evaluate a real-time flaw detection and characterization system based on the synthetic aperture focusing technique for ultrasonic testing (SAFT-UT). The system is designed to perform inservice inspection of light-water reactor components. Included objectives of this program for the Nuclear Regulatory Commission are to develop procedures for system calibration and field operation, to validate the system through laboratory and field inspections, and to generate an engineering database to support ASME Code acceptance of the technology. This progress report covers the programmatic work from October 1986 through September 1987. (author)

Implementation of in-service inspection program for HANARO

International Nuclear Information System (INIS)

Wu, J.S.; Park, Y.C.; Cho, Y.G.; Jun, B.J.

2001-01-01

HANARO, a 30 MW multi-purpose research reactor in Korea has been successfully in operation for 6 years since its initial criticality in February 1995. It is mainly used for the research areas including nuclear fuel and material irradiation tests, radioisotope production, neutron beam application, neutron activation analysis and neutron transmutation doping. HANARO was designed to perform for at least 20 years under full power operating condition. It is expected that the actual reactor lifetime will be much more than the design lifetime, due to a safety reassessment based on realistic data, preventive maintenance and appropriate in-service inspections (ISI). Since ageing may affect the overall safety of the reactor facility, it is needed to detect and evaluate the effects on aged components and systems related to safety. During the lifetime of the reactor, structures, systems and components are subjected to environmental conditions of stress, temperature and irradiation that may lead to changes in the material properties and could result in unexpected failures. Evidence of ageing problems appears progressively. A rigorous inspection and visual examination based on a periodic ISI program should be established. It is desirable that the ageing surveillance activities is scheduled as early as possible and continued throughout the operating life of the reactor. An inspection plan for safety related structures, systems and components subjected to the ageing conditions is requested by the regulatory body to assess the safety status of reactor facility. A long-term ISI program for HANARO has been established for safety-related systems and components in the context of the overall reactor ageing management. The objective of this paper is to describe the ISI program and the result of the visual inspection as the first ISI. (orig.)

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)

Evaluation of computer-based ultrasonic inservice inspection systems

International Nuclear Information System (INIS)

Harris, R.V. Jr.; Angel, L.J.; Doctor, S.R.; Park, W.R.; Schuster, G.J.; Taylor, T.T.

1994-03-01

This report presents the principles, practices, terminology, and technology of computer-based ultrasonic testing for inservice inspection (UT/ISI) of nuclear power plants, with extensive use of drawings, diagrams, and LTT images. The presentation is technical but assumes limited specific knowledge of ultrasonics or computers. The report is divided into 9 sections covering conventional LTT, computer-based LTT, and evaluation methodology. Conventional LTT topics include coordinate axes, scanning, instrument operation, RF and video signals, and A-, B-, and C-scans. Computer-based topics include sampling, digitization, signal analysis, image presentation, SAFI, ultrasonic holography, transducer arrays, and data interpretation. An evaluation methodology for computer-based LTT/ISI systems is presented, including questions, detailed procedures, and test block designs. Brief evaluations of several computer-based LTT/ISI systems are given; supplementary volumes will provide detailed evaluations of selected systems

Development of Pre-Service and In-Service Information Management System (iSIMS)

International Nuclear Information System (INIS)

Yoo, H. J.; Choi, S. N.; Kim, H. N.; Kim, Y. H.; Yang, S. H.

2004-01-01

The iSTMS is a web-based integrated information system supporting Pre-Service and In-Service Inspection(PSI/ISI) processes for the nuclear power plants of KHNP(Korea Hydro and Nuclear Power Co. Ltd.). The system provides a full spectrum coverage of the inspection processes from the planning stage to the final report of examination in accordance with applicable codes, standards, and regulatory requirements. The major functions of the system includes the inspection planning, examination, reporting, project control and status reporting, resource management as well as objects search and navigation. The system also provides two dimensional or three dimensional visualization interface to identify the location and geometry of components and weld areas subject to examination in collaboration with database applications. The iSIMS is implemented with commercial software packages such as database management system, 2-D and 3-D visualization tool, etc., which provide open, updated and verified foundations. This paper describes the key functions and the technologies for the implementation of the iSIMS

Development of Pre-Service and In-Service Information Management System (iSIMS)

Energy Technology Data Exchange (ETDEWEB)

Yoo, H. J.; Choi, S. N.; Kim, H. N.; Kim, Y. H.; Yang, S. H. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2004-08-15

The iSTMS is a web-based integrated information system supporting Pre-Service and In-Service Inspection(PSI/ISI) processes for the nuclear power plants of KHNP(Korea Hydro and Nuclear Power Co. Ltd.). The system provides a full spectrum coverage of the inspection processes from the planning stage to the final report of examination in accordance with applicable codes, standards, and regulatory requirements. The major functions of the system includes the inspection planning, examination, reporting, project control and status reporting, resource management as well as objects search and navigation. The system also provides two dimensional or three dimensional visualization interface to identify the location and geometry of components and weld areas subject to examination in collaboration with database applications. The iSIMS is implemented with commercial software packages such as database management system, 2-D and 3-D visualization tool, etc., which provide open, updated and verified foundations. This paper describes the key functions and the technologies for the implementation of the iSIMS

Ultrasonic wall thickness gauging for ferritic steam generator tubing as an in-service inspection tool

International Nuclear Information System (INIS)

Haesen, W.M.J.; Tromp, Th.J.

1980-01-01

In-service inspection of LWR steam generators is more or less a standard routine operation. The situation can be very different for LMFBRs. For the SNR 300 (Kalkar Power Station) the situation is different because the steam generators have ferritic tubing. The tube walls are comparatively thick, 2 to 4.5 mm. During inservice examinations the steam generators will be drained on both sides, however on the sodium side a sodium film will be present. Furthermore the SNR 300 will have two types of steam generator. A straight tube design and a helical coil design will be used. Both types consist of a evaporator and superheater. The steam generators are of course not radioactive. It is obvious that in this case the eddy current (EC) technique is not an enviable inservice inspection tool. Basically EC is a surface flaw detection technique. Only the saturation magnetisation method will improve the EC technique sufficiently for ferritic material. However the 'in bore examination' with the saturation technique was, in case of the SNR 300 steam generator tubing, considered impossible since the inner diameters are fairly small. Furthermore sodium traces may influence the EC method. Although multifrequency methods can solve this problem, EC is not considered as a useful tool for examining ferritic tubing. Another method is to employ the 'stray flux' method which is under development with the TNO organization in Holland. The EC and stray flux method do have one drawback, these methods do not detect gradual changes in wall thickness. Ultrasonic examinations will be used in the SNR 300 as the main inspection tool for the steam generators. In this paper the reasons why ultrasonic examination was selected are explained. The results of the development work on this subject are discussed

RSE-M: In-Service Inspection Rules for Mechanical Components of PWR Nuclear Islands

International Nuclear Information System (INIS)

2016-01-01

The RSE-M code defines in-service inspection operations. It applies to pressure equipment used in PWR plants, as well as spare parts for such equipment. The RSE-M code does not apply to equipment made from materials other than metal. It is based on the RCC-M code for requirements relating to the design and fabrication of mechanical components. Use: The inspection rules specified in the RSE-M code describe the standard requirements of best practice within the French nuclear industry, based on its own feedback from operating several nuclear units and partly supplemented with requirements stipulated by French regulations. To date, the 58 units in France's nuclear infrastructure enforce the in-service inspection rules of the RSE-M code. Operation of 30 commissioned units in China's nuclear infrastructure, corresponding to the M310, CPR-1000 and CPR-600 reactors, is based on the RSE-M code (since 2007, use of AFCEN codes has been required by NNSA for Generation II+ reactors). Contents of the 2016 Edition: Volume I - Rules: Section A - General rules, Section B - Specific rules for class 1 components, Section C - Specific rules for class 2 or 3 components, Section D - Specific rules for components not assigned to any particular RSE-M class; Volume II - Appendices 1 to 8: Appendices 1.0 to 1.9: supporting appendices for the general requirements, Appendix 2.1: appendix associated with chap. 2000 Requalifications, Hydraulic Proof Tests and Hydraulic Tests, Appendices 4.1 to 4.4: appendices associated with chap. 4000 Examination techniques, Appendices 5.1 to 5.8 and RPP2: appendices associated with chap. 5000 Mechanical and Materials, Appendices 8.1 to 8.2: appendices associated with chap. 8000 Maintenance Operations; Volume III: Appendix 3.1 - Visit tables: main primary and secondary systems, EPR pre-service inspection program, Class 2 or 3 vessels; Appendix 3.2 - Inspection Plans For Non-Nuclear Pressure Equipment

Risk-informed optimisation of railway tracks inspection and maintenance procedures

International Nuclear Information System (INIS)

Podofillini, Luca; Zio, Enrico; Vatn, Jorn

2006-01-01

Nowadays, efforts are being made by the railway industry for the application of reliability-based and risk-informed approaches to maintenance optimisation of railway infrastructures, with the aim of reducing the operation and maintenance expenditures while still assuring high safety standards. In particular, in this paper, we address the use of ultrasonic inspection cars and develop a methodology for the determination of an optimal strategy for their use. A model is developed to calculate the risks and costs associated with an inspection strategy, giving credit to the realistic issues of the rail failure process and including the actual inspection and maintenance procedures followed by the railway company. A multi-objective optimisation viewpoint is adopted in an effort to optimise inspection and maintenance procedures with respect to both economical and safety-related aspects. More precisely, the objective functions here considered are such to drive the search towards solutions characterized by low expenditures and low derailment probability. The optimisation is performed by means of a genetic algorithm. The work has been carried out within a study of the Norwegian National Rail Administration (Jernbaneverket)

Optimizing pre-service inspection data

International Nuclear Information System (INIS)

Mullins, L.E.; Harrison, J.M.

1986-01-01

The phenomenon of intergranular stress corrosion cracking (IGSCC) in nuclear power plant piping systems has been the single biggest factor in an evolution that has changed and continues to change in-service inspection practices and philosophy throughout the industry. In-service inspection has been found to deal with several distinct phases of the problem. At the onset, existing technology was demonstrated to be unable to detect IGSCC adequately. This inability to detect IGSCC was soon overcome. However, it immediately pointed out the difficulties associated with the discrimination between IGSCC and other harmless metallurgical conditions. On the heels of detection and discrimination came problems associated with determining crack size and the monitoring of overlayed flaws. A wealth of information has been written on all phases of IGSCC; however, one particular aspect has as of yet seen little attention. With an increasing number of plants choosing to replace flawed pipe, the pre-service inspection of the replaced piping demands increased attention. The lessons learned through inspection for IGSCC need again to be applied to the replacement piping. Through increased attention at installation, many of the difficulties associated with proper inservice inspection may be reduced or eliminated, thereby realizing an economy in outage scheduling, personnel exposure, and, overall, a more reliable in-service examination

Pre-service inspection and in-service inspection in Japan

International Nuclear Information System (INIS)

Uebayashi, T.; Miyake, Y.

1985-01-01

To ensure the safety of nuclear power plant, pre-service inspection/in-service inspection (PSI/ISI) has an important role, and informations obtained from various inspections during plant shut-down period are contributing to establish effective preventive maintenance activities for plant facilities. It might be said that the high level of availability of Japanese light-water nuclear power plants in these two or three years has been achieved by those efforts. In case of Japan, inspections to be carried out during scheduled plant shut-down period are not limited to code requirements but include many other inspections which are mostly reflected from troubles experienced in both domestic and overseas plants. Usually, those additional inspections are performed by Ministry of Trade and Industries' (MITI's) regulator and/or tentative requirement and considered as ISI in broad meaning. To achieve high availability of plant, it is essential to avoid unscheduled shut-down and to shorten inspection period. The developments of new technology to perform effective ISI for operating plants are continued, but on the other hand it is also very important to pay a great consideration to inspectability of the plants at the stage of plant engineering. With the leadership of MITI, improvement and standardization of light-water nuclear power plant has been proceeded and newly constructed plants have great advantage from the point of view on ISI

Development, design, construction and testing of pumps, intermediate heat exchangers and manipulator systems for in-service inspection of SNR 300

International Nuclear Information System (INIS)

Vroom, J.P.; Heslenfeld, M.W.; Westerweele, W.J.; Willigen, A. van

1978-01-01

The first part of this paper reviews the development, design, construction and testing of sodium pumps and intermediate heat exchangers as carried out by NERATOOM in the Netherlands. Testing of full-size components under operational conditions yielded important information for the design of the components for SNR 300. Information is also given in brief on the manufacture of the components. The second part reviews the work done on manipulator systems for in-service inspection as carried out by INTERATOM in Germany. In this field again experience gained from tests had important consequences for the design of the apparatus for SNR 300. (author)

Review of P-scan computer-based ultrasonic inservice inspection system. Supplement 1

International Nuclear Information System (INIS)

Harris, R.V. Jr.; Angel, L.J.

1995-12-01

This Supplement reviews the P-scan system, a computer-based ultrasonic system used for inservice inspection of piping and other components in nuclear power plants. The Supplement was prepared using the methodology described in detail in Appendix A of NUREG/CR-5985, and is based on one month of using the system in a laboratory. This Supplement describes and characterizes: computer system, ultrasonic components, and mechanical components; scanning, detection, digitizing, imaging, data interpretation, operator interaction, data handling, and record-keeping. It includes a general description, a review checklist, and detailed results of all tests performed

Risk-informed regulation

International Nuclear Information System (INIS)

Hoffman, D.R.

2003-01-01

In assessing safety for nuclear facilities, regulators have traditionally used a deterministic approach. New techniques for assessing nuclear or radiological risks make it possible for regulators to incorporate risk insights into their regulations. By 'risk-informing' the regulatory processes, independent bodies tasked with protecting the health and safety of the public can focus on those design and operational issues most important to safety. Such an approach is a move away from prescriptive regulations that were based on conservative engineering judgments toward regulations focused on issues that contribute significantly to safety. Despite the availability of probabilistic risk assessment (PRA) tools, organisations often struggle with how to best use this capability. Most international regulations are still based largely on deterministic analyses that were developed without the benefit of quantitative or measurable estimates of risk. PRA considers issues of risk in a more comprehensive manner by examining a wider spectrum of initiating events and their frequency, and considers the likelihood of events in a rigorous and comprehensive manner. In some countries, nuclear regulators are actively moving toward increasing the use of risk insights in a variety of strategic arenas, including risk-informed technical specifications (operating limits and conditions), in-service inspection and testing, programs, and assessment and enforcement actions. A risk-informed approach enhances the traditional deterministic approach by explicitly considering a broader range of safety challenges, focusing resources on the basis of risk significance, considering a broader range of counter measures to mitigate challenges, and explicitly identifying and quantifying uncertainties in analyses. (author)

Studies on the effect of flaw detection probability assumptions on risk reduction at inspection

Energy Technology Data Exchange (ETDEWEB)

Simola, K.; Cronvall, O.; Maennistoe, I. (VTT Technical Research Centre of Finland (Finland)); Gunnars, J.; Alverlind, L.; Dillstroem, P. (Inspecta Technology, Stockholm (Sweden)); Gandossi, L. (European Commission Joint Research Centre, Brussels (Belgium))

2009-12-15

The aim of the project was to study the effect of POD assumptions on failure probability using structural reliability models. The main interest was to investigate whether it is justifiable to use a simplified POD curve e.g. in risk-informed in-service inspection (RI-ISI) studies. The results of the study indicate that the use of a simplified POD curve could be justifiable in RI-ISI applications. Another aim was to compare various structural reliability calculation approaches for a set of cases. Through benchmarking one can identify differences and similarities between modelling approaches, and provide added confidence on models and identify development needs. Comparing the leakage probabilities calculated by different approaches at the end of plant lifetime (60 years) shows that the results are very similar when inspections are not accounted for. However, when inspections are taken into account the predicted order of magnitude differs. Further studies would be needed to investigate the reasons for the differences. Development needs and plans for the benchmarked structural reliability models are discussed. (author)

Studies on the effect of flaw detection probability assumptions on risk reduction at inspection

International Nuclear Information System (INIS)

Simola, K.; Cronvall, O.; Maennistoe, I.; Gunnars, J.; Alverlind, L.; Dillstroem, P.; Gandossi, L.

2009-12-01

The aim of the project was to study the effect of POD assumptions on failure probability using structural reliability models. The main interest was to investigate whether it is justifiable to use a simplified POD curve e.g. in risk-informed in-service inspection (RI-ISI) studies. The results of the study indicate that the use of a simplified POD curve could be justifiable in RI-ISI applications. Another aim was to compare various structural reliability calculation approaches for a set of cases. Through benchmarking one can identify differences and similarities between modelling approaches, and provide added confidence on models and identify development needs. Comparing the leakage probabilities calculated by different approaches at the end of plant lifetime (60 years) shows that the results are very similar when inspections are not accounted for. However, when inspections are taken into account the predicted order of magnitude differs. Further studies would be needed to investigate the reasons for the differences. Development needs and plans for the benchmarked structural reliability models are discussed. (author)

ASME section XI - design and access requirements for in-service inspection

International Nuclear Information System (INIS)

Davis, D.D.

1982-01-01

The Owner of a nuclear power plant has the regulatory commitment to perform Section XI in-service inspection throughout the service life of a plant. In anticipation of what will be needed to perform adequately the required examinations and tests, sub-article IWA-1500 of Section XI not only requires that sufficient access be provided to accommodate equipment and inspection personnel but also requires that other provisions be considered such as: component surface preparations, material selections, shielding, removal and storage of hardware, handling equipment, and provisions for repairs and replacements. It is, therefore, the owner's and the architect engineer's responsibility to ensure that proper design and access provisions are incorporated to enable the owner to meet his commitments. Since the architect engineer usually has the prime responsibility for the implementation of design criteria, the owner must ensure that these provisions be considered in each phase of design and construction. The benefits of this can result in shorter outages, more meaningful examinations and tests and less radiation exposure of inspection personnel. This paper will address in detail those topics that affect design and access provisions which need to be considered during the design and construction of a nuclear power plant. (author)

Qualification of NDT systems for in-service inspections of nuclear power plant pressure vessels

International Nuclear Information System (INIS)

Elfving, K.

1998-11-01

The goal of this study is to determine the requirements of the in-service inspection qualification in Europe, their feasibility in practice and to find out possible manufacture defects in test pieces used in practical trials. The literature study consists of qualification requirements set by European regulatory bodies and by the European nuclear power utilities. Also a brief summary of qualification requirements set by ASME Code, Section XI and comparison between ASME and European qualification requirements is included

X-ray tomographic in-service inspection of girth welds - The European project TomoWELD

International Nuclear Information System (INIS)

Ewert, Uwe; Redmer, Bernhard; Walter, David; Thiessenhusen, Kai-Uwe; Bellon, Carsten; Nicholson, P. Ian; Clarke, Alan L.; Finke-Härkönen, Klaus-Peter

2015-01-01

The new standard 'ISO 17636-2:2013: Non-destructive testing of welds - Radiographic testing - Part 2: X- and gamma-ray techniques with digital detectors', defines the practice for radiographic inspection of welded pipes for manufacturing and in-service inspection. It is applied in Europe for inspections of pipe welds in nuclear power plants as well as in chemical plants and allows a faster inspection with digital detector arrays (DDA) than with film. Nevertheless, it does not allow the evaluation of the depth and shape of volumetric and planar indications. In 2001 a planar tomography scanner, TomoCAR, was introduced for mechanized radiographic testing (RT) inspection and non-destructive measurement of cross sections. The project TomoWELD is based on a new concept of the scan geometry, an enhanced GPU based reconstruction, and the application of a new generation of photon counting DDAs based on CdTe crystal CMOS hybrids. The new detector permits the selection of energy thresholds to obtain an optimum energy range and reduction of the influence of scattered radiation. The concept and first measurements are presented. Flaw depth and shape of volumetric and planar irregularities can be determined

X-ray tomographic in-service inspection of girth welds - The European project TomoWELD

Energy Technology Data Exchange (ETDEWEB)

Ewert, Uwe, E-mail: uwe.ewert@bam.de; Redmer, Bernhard, E-mail: uwe.ewert@bam.de; Walter, David, E-mail: uwe.ewert@bam.de; Thiessenhusen, Kai-Uwe, E-mail: uwe.ewert@bam.de; Bellon, Carsten, E-mail: uwe.ewert@bam.de [BAM Bundesanstalt für Materialforschung, Berlin (Germany); Nicholson, P. Ian, E-mail: ian.nicholson@twi.co.uk [TWI Technology Centre Wales, Port Talbot (United Kingdom); Clarke, Alan L. [TWI Technology Centre Wales, Port Talbot (United Kingdom); Finke-Härkönen, Klaus-Peter, E-mail: klaus.harkonen@ajat.fi [Oy AJAT Ltd., Helsinki (Finland)

2015-03-31

The new standard 'ISO 17636-2:2013: Non-destructive testing of welds - Radiographic testing - Part 2: X- and gamma-ray techniques with digital detectors', defines the practice for radiographic inspection of welded pipes for manufacturing and in-service inspection. It is applied in Europe for inspections of pipe welds in nuclear power plants as well as in chemical plants and allows a faster inspection with digital detector arrays (DDA) than with film. Nevertheless, it does not allow the evaluation of the depth and shape of volumetric and planar indications. In 2001 a planar tomography scanner, TomoCAR, was introduced for mechanized radiographic testing (RT) inspection and non-destructive measurement of cross sections. The project TomoWELD is based on a new concept of the scan geometry, an enhanced GPU based reconstruction, and the application of a new generation of photon counting DDAs based on CdTe crystal CMOS hybrids. The new detector permits the selection of energy thresholds to obtain an optimum energy range and reduction of the influence of scattered radiation. The concept and first measurements are presented. Flaw depth and shape of volumetric and planar irregularities can be determined.

Activities with regard to research and development of technics for SNR 300 reactor vessel in-service inspection procedures

International Nuclear Information System (INIS)

Hoeller, K.; Kirchner, G.; Menck, J.

1980-01-01

During the development of SNR 300 in-service inspection equipment, several branches were tested by experiment. In this report special steps for testing of manipulation systems and additional engineering equipment for control systems, such as coupling fluid circuit or camera cooling system are considered more in detail

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)

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

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

Some reliability targets affecting the necessary provisions for in-service inspection and monitoring of LMFBR engineering components

International Nuclear Information System (INIS)

Bolt, P.R.

1980-01-01

The possible consequences of failure of primary and secondary sodium circuit components are discussed with particular reference to post incident fault diagnosis, remedial procedures and outage durations. The core support structures and steam generator units are identified as particularly important components in terms of economic consequence of their failure. Important safety considerations may also apply. Levels of reliability for core support and steam generator integrity, necessary to meet economic and certain safety criteria, are discussed and quantitative data is given. Possible failure and deterioration mechanisms which could result in unacceptable reductions in reliability are then identified for the core support and steam generator units. Following a consideration of the reliability targets and possible causes of loss of reliability, an appraisal is made of the necessary extent of in-service data to be obtained on component behaviour and condition. In-service inspection and monitoring methods that could be used to obtain this data are described. Consideration is given to UK and overseas inspection experience on LMFBR and other nuclear plant. (author)

Some reliability targets affecting the necessary provisions for in-service inspection and monitoring of LMFBR engineering components

Energy Technology Data Exchange (ETDEWEB)

Bolt, P R [Fast Reactor Engineering, Plant Engineering Department, CEGB, Barnwood, Gloucester (United Kingdom)

1980-11-01

The possible consequences of failure of primary and secondary sodium circuit components are discussed with particular reference to post incident fault diagnosis, remedial procedures and outage durations. The core support structures and steam generator units are identified as particularly important components in terms of economic consequence of their failure. Important safety considerations may also apply. Levels of reliability for core support and steam generator integrity, necessary to meet economic and certain safety criteria, are discussed and quantitative data is given. Possible failure and deterioration mechanisms which could result in unacceptable reductions in reliability are then identified for the core support and steam generator units. Following a consideration of the reliability targets and possible causes of loss of reliability, an appraisal is made of the necessary extent of in-service data to be obtained on component behaviour and condition. In-service inspection and monitoring methods that could be used to obtain this data are described. Consideration is given to UK and overseas inspection experience on LMFBR and other nuclear plant. (author)

Surveillance and in-service inspection (SISI) program at FFTF

International Nuclear Information System (INIS)

Conrads, T.J.

1980-01-01

Assurance of the integrity of the sodium coolant pressure boundaries of the Fast Flux Test Facility systems and components is essential for safe operation. A program has been developed to monitor the integrity of the coolant boundaries and certain plant conditions. Specific equipment and features have been designed into the plant for monitoring. The purpose of SISI is to prevent failures or minimize their consequences through early detection. The program which administers the requirements for monitoring applicable plant conditions whose integrity is necessary to protect public health and safety is known as the Surveillance and In-service Inspection (SISI) Program. The SISI program utilizes a wide range of monitoring techniques to ensure that material degradation or structural deficiencies will not result in the loss of the ability to shut down the reactor, cool the reactor core, or cause the release of radioactive material to the environment

Evaluation and improvement of nondestructive evaluation reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Bates, D.J.; Deffenbaugh, J.D.; Good, M.S.; Heasler, P.G.; Mart, G.A.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Van Fleet, L.G.

1987-01-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection (ISI) of Light Water Reactors (NDE Reliability) Program at Pacific Northwest Laboratory (PNL) was established to determine the reliability of current ISI techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this NRC program are to: determine the reliability of ultrasonic ISI performed on commercial light-water reactor (LWR) primary systems, using probabilistic fracture mechanics analysis, determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to ensure a suitably low failure probability, evaluate the degree of reliability improvement that could be achieved using improved and advanced NDE techniques, based on material properties, service conditions, and NDE uncertainties, recommend revisions to ASME Code, Section XI, and Regulatory Requirements that will ensure suitably low failure probabilities. The scope of this program is limited to ISI of primary systems; the results and recommendations may also be applicable to Class II piping systems

Careful determination of inservice inspection of piping by computer analysis in nuclear power plant

International Nuclear Information System (INIS)

Lim, H. T.; Lee, S. L.; Lee, J. P.; Kim, B. C.

1992-01-01

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in order to predict possibility of crack generation due to thermal stratification phenomena in pipes connected to reactor coolant system of Nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.

Risk-informed importance analysis of in-service testing components for Ulchin units 3 and 4

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.; Ha, J. J.

2001-01-01

In this paper, we perform risk-informed importance analysis of in-service tesing (IST) components for Ulchin Units 3 and 4. The importance analysis using PSA is performed through Level 1 internal and external, shutdown/low power operation, and Level 2 internal PSA. The sensitivity analysis is also performed. For the components not modeled in PSA logic, we develop and apply a new integrated importance analysis method. The importance analysis results for IST valves show that 167 (26.55%) of 629 IST valves are HSSCs and 462 (73.45%) are LSSCs. The importance analysis results for IST pumps show that 28 (70%) of 40 IST pumps are HSSCs and 12 (30%) are KSSCs

In-Service Science Teachers' Attitude towards Information Communication Technology

Science.gov (United States)

Kibirige, I.

2011-01-01

The purpose of this study is to determine the attitude of in-service science teachers towards information communication technology (ICT) in education. The study explores the relationship between in-service teachers and four independent variables: their attitudes toward computers; their cultural perception of computers; their perceived computer…

Careful Determination of Inservice Inspection of piping by Computer Analysis in Nuclear Power Plant

International Nuclear Information System (INIS)

Lim, H. T.; Lee, S. L.; Lee, J. P.; Kim, B. C.

1992-01-01

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants

Design considerations, tooling and equipment for remote in-service inspection of radioactive piping and pressure vessel systems

International Nuclear Information System (INIS)

Schmoker, D.S.; Swannack, D.L.

1983-01-01

In-Service Inspection programs are performed to monitor and verify the integrity of a nuclear power plant's primary pressure boundaries. Early detection of abnormal structural or material degradation could preclude serious damage to plant systems. This paper summarizes results obtained in use of remotely-operated nondestructive testing (NDT) equipment for inspection of reactor system components. Experience obtained in operating the Fast Flux Test Facility (FFTF) has provided a basis for field verification of remote NDT equipment designs and has suggested development improvements. Remote Viewing and data gathering systems used include periscopes, borescopes, fiberscopes, hybrid borescopes/fiberscopes, and closed circuit television. A summary of design consideration for inspection equipment and power plant design is presented to achieve improved equipment operation and reduction of plant maintenance downtime

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

Nuclear reactor capable of electric power generation during in-service inspection

International Nuclear Information System (INIS)

Nakamura, Shinsuke; Nogami, Hitoshi.

1992-01-01

The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

Risk analysis of in-service pressure piping containing defects

International Nuclear Information System (INIS)

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

2004-01-01

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

Example value-impact analysis of non-destructive examination methods used for inservice inspection of BWR piping

International Nuclear Information System (INIS)

Tabatabai, A.S.; Simonen, F.A.

1985-12-01

This paper describes work recently completed at Pacific Northwest Laboratory (PNL) to use value-impact (V/I) analysis methods to help guide research to improve the effectiveness of inservice inspection (ISI) procedures at nuclear power plants. The example developed at PNL uses the results of probabilistic fracture mechanics and probabilistic risk analysis (PRA) studies to compare three generic categories of non-destructive examination (NDE) methods. These NDE methods are used to detect possible pipe cracks such as those induced by intergranular stress corrosion (IGSCC). The results of the analysis of this example include (1) quantification of the effectiveness of ISI in increasing plant safety in terms of reduction in core-melt frequency, (2) estimates of the industry cost of performing ISI, (3) estimates of radiation exposures to plant personnel as a result of performing ISI, and (4) potential areas of improvement in the NDE and ISI process

Periodical in-service inspection as part of individual program of quality assurance of steam generators and pressurizers of WWER 440 nuclear power plant

International Nuclear Information System (INIS)

Kawalec, M.

1982-01-01

The manufacturers of equipment for nuclear power plants in the Czechoslovak Socialist Republic are obligated to process so-called individual programs of quality assurance in order to secure the quality of selected equipment in nuclear power. These programmes should include the evaluation of the design of the individual equipments with regard to the implementation of in-service inspection. The main problems are discussed related to the processing of the program of quality assurance for the steam generator and pressurizer. To solve these problems it is necessary that the general project designer should make a classification of the components according to safety categories and that the manufacturers should determine the weak points of the design on the basis of an analysis of the design of individual component nodes. On the basis of such an analysis it is then necessary to evaluate the existing design of the scale of in-service inspections and to decide whether or not new inspection methods should be added. (Z.M.)

Developments on ASME Code Cases to Risk-Informed Repair/Replacement Activities in Support of Risk-Informed Regulation Initiatives

International Nuclear Information System (INIS)

Balkey, Kenneth R.; Holston, William C.

2002-01-01

ASME Code Case N-658, 'Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities' and Code Case N-660, 'Alternative Repair/Replacement Requirements For Items Classified In Accordance With Risk-Informed Processes' are being completed to expand the breadth of risk-informed requirements for pressure-retaining items. This initiative, which is built from prior ASME Section XI risk-informed inservice inspection developments over the past decade, has been undertaken in conjunction with U.S. risk-informed regulation efforts. The U.S. Nuclear Regulatory Commission (NRC) is working with the industry on risk informing Title 10 Code of Federal Regulations Part 50 (10CFR50). The Nuclear Regulatory Commission's basic proposal is to allow modification of some of the special treatment requirements of 10CFR50. Their effort is proceeding via an Advanced Notice of Public Rulemaking, March 3, 2000, and an announcement of Availability of Draft Rule Wording, November 29, 2001, to add 10 CFR 50.69, 'Risk-Informed Treatment of Structures, Systems and Components'. A parallel task by the Nuclear Energy Institute (NEI) to develop a guideline on how to implement the results of the rulemaking is also well underway via NEI 00-04 (Draft Revision B), 'Option 2 Implementation Guideline', May 2001. This paper summarizes the content and status of approval of the proposed ASME Code Cases, including how they relate to the above NRC and NEI efforts. Some initial results from trial application of the Code Cases will also be cited. (authors)

Pre- and in-Service Teachers Reading and Discussing Informational Texts

Directory of Open Access Journals (Sweden)

Theresa A. Deeney

2016-05-01

Full Text Available This study investigates U.S. elementary (kindergarten-Grade 6, ages 5-12 pre- and in-service teachers’ discussions of informational texts to understand current practices and identify needs with respect to how teachers support students in building knowledge from complex informational text as specified in the grade-level expectations of the Common Core State Standards adopted in many U.S. states. Transcripts and reflections from 17 in-service and 31 pre-service teachers’ informational text discussions were analyzed for teachers’ focus on the text, background knowledge, and text/background knowledge. In addition, transcripts were analyzed for the types of text ideas teachers targeted (details/main ideas, the comprehension demands placed on students, how teachers used follow-up moves to encourage higher level thinking, and how teachers use transcripts of their discussions to analyze and critique their own practice. Findings suggest that both pre- and in-service teachers draw heavily on students’ background knowledge and text details in their questioning; but differences exist in how pre- and in-service teachers use follow-up responses to promote knowledge building. Findings also suggest that both pre- and in-service teachers can use their transcripts to recognize areas of need, and offer themselves suggestions to better support students’ understanding. Implications are offered for teacher education and professional development.

An in-service inspection method: the use of focused probes for the detection and sizing in DDT Plates 1 and 2

International Nuclear Information System (INIS)

Birac, A.M.; Pincemaille, G.; Saglio, R.; Cattiauz, G.; Morisseau, P.

1983-01-01

As part of the DDT programme to evaluate the nondestructive testing methods used during the in-service inspection (ISI) of PWR vessels, the UKAEA asked the CEA-STA-SCND to perform the ultrasonic examination of two test-plates. Each plate was bisected across its full width by a full-thickness butt-weld in which calibrated defects of various natures, dimensions, orientations have been inserted. It was required that the implementation of control, the equipment and the subsequent data analysis represent as nearly as possible the French procedure in force during the pre-service and in-service inspection of a PWR vessel. As suggested by the UKAEA, to evaluate the influence of examiners on the results, the examinations have been carried out twice by two independent teams: one from the CEA-STA-SCND and the second from INTERCONTROLE (a CEA subsidiary company). Procedure and results are given. (author)

Recent experiences with ultrasonic inservice inspection systems with phased array probes on spherical bottoms of boiling water reactors

International Nuclear Information System (INIS)

Wustenberg, H.; Brekow, G.; Erhard, A.; Hein, E.

1988-01-01

The special geometry of the spherical bottom of boiling water reactors with control rods and measuring nozzles requires a very special surveillance technique during the in-service inspection. Reside visual inspection an ultrasonic inspection has been established due to the requirements of German authorities. A first application of a new phased array system took place August 1987. The 100% inspection of a spherical bottom had been enabled by the application of phased array probes with electronically controlled skewing angles. The data acquisition had been based on the storage of whole A-scans, which had been pixellized into 256 points. This A-scan storage procedure makes possible the application of a simple and fast algorithm to present the data as TD-(time displacement)-scans. Defect reconstruction by echotomographique approaches are under development. This paper presents the ultrasonic technique applied including the phased array probes, the electronic system, as well as the software package used for the control of the inspection parameters depending on the probe position

Reactor safety through quality assurance and in-service inspection

International Nuclear Information System (INIS)

Bush, S.H.

The quality assurance is discussed of nuclear power plant equipment with respect to the following regulations: section 50 10 CFR - supplement B, section NA-400, ASME - section III and ANSI N-54.2. Quality assurance and reliability are assessed with regard to two aspects: all preoperational functions and all operating stages of the power plant. During the production of nuclear power plant components, increased attention should be devoted to the choice of material, materials testing, production programme and to the production process. During power plant operation, care should be given to periodical in-service inspections which guarantee the plant reliability; defects should immediately be remedied or the power plant shut down. Emphasis is put on the tests of reactor welded joints in compliance with the ASME code. The results of operating tests are used as feedback in the design and testing of the components during production. The probabilities were calculated of the occurrence and elimination of defects during the manufacture and operation of a nuclear reactor. (J.B.)

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.

Risk evaluation for motor operated valves in an Inservice Testing Program at a PWR nuclear power plant in Taiwan

International Nuclear Information System (INIS)

Li, Y.C.; Chen, K.T.; Su, Y.L.; Ting, K.; Chien, F.T.; Li, G.D.; Huang, S.H.

2012-01-01

Safety related valves such as Motor Operated Valves (MOV), Air Operated Valves (AOV) or Check Valves (CV) play an important role in nuclear power plant. Functioning of these valves mainly aim at emergency reactivity control, post-accident residue heat removal, post-accident radioactivity removal and containment isolation when a design basis accident occurred. In order to maintain these valves under operable conditions, an Inservice Testing Program (IST) is defined for routine testing tasks based on the ASME Boiler and Pressure Vessel Code section XI code requirements. Risk based Inservice Testing Programs have been studied and developed extensively in the nuclear energy industry since the 1990s. Risk Based evaluations of IST can bring positive advantages to the licensee such as identifying the vulnerability of the system, reducing unnecessary testing burden, concentrating testing resources on the critical pass oriented valves and saving plant’s personnel dose exposure. This risk evaluation is incorporated with quantitative and qualitative analyses to the Motor Operated Valves under current Inservice Testing Program for PWR nuclear power plant in Taiwan. With the outcome of the risk classifications for the safety related MOVs through numerical or deterministic analyses, a risk based testing frequency relief is suggested to demonstrate the benefits received from the risk based Inservice Testing Program. The goal made of this study, it could be as a reference and cornerstone for the licensee to perform overall scope Risk-Informed Inservice Testing Program (RI-IST) evaluation by referring relevant methodologies established in this study.

In-Service Inspection system for coolant channels of Indian PHWRS - evolution and experience

International Nuclear Information System (INIS)

Puri, R.K.; Singh, M.

2006-01-01

In-Service Inspection (ISI) is the most important of all periodic monitoring and surveillance activities for assuring the structural integrity of coolant channels in the life extension and management of pressurized heavy water reactors (PHWR-CANDU). Indian PHWRs (220 MWe) are characterized by consists by 306 coolant channels in each unit. These channels have to be inspected for various parameters over the operating life of the reactor. ISI of coolant channels necessitated the indigenous development of an inspection system called BARCIS (BARC Channel Inspection System) at Bhabha Atomic Research Center. BARCIS consists of mainly three parts; drive and control unit, special sealing plug and an inspection head carrying various NDT sensors. Five such systems have been built and deployed at various power plants. The paper deals with the development of the BARCIS system for meeting the ISI requirements of coolant channels, development cycle of this system from its conception to evolution to the present state, challenges, data generated and experience gained (ISI of nearly 900 coolant channels has been completed). Prior to BARCIS, pressure tube gauging equipment for pre-service inspection of coolant tubes was developed in 1980. Moreover a tool for ISI of coolant channels in dry condition was developed in 1990. The paper also describes evolution of various contingency procedures and devices developed over the last one decade. Future plans taking into account technological advancement, changes in the scope of inspection due to design and operating experiences and plant layout will also be covered. The paper describes the efforts put in to develop drive and control mechanism to suit the different vault layouts. The drive mechanism is responsible for linear and rotary movement of the inspection head to carry out 100% volumetric inspection. Special emphasis has been laid on the safety devices required during the inspection activity. Special measures for heavy water retention in

Current food chain information provides insufficient information for modern meat inspection of pigs.

Science.gov (United States)

Felin, Elina; Jukola, Elias; Raulo, Saara; Heinonen, Jaakko; Fredriksson-Ahomaa, Maria

2016-05-01

Meat inspection now incorporates a more risk-based approach for protecting human health against meat-borne biological hazards. Official post-mortem meat inspection of pigs has shifted to visual meat inspection. The official veterinarian decides on additional post-mortem inspection procedures, such as incisions and palpations. The decision is based on declarations in the food chain information (FCI), ante-mortem inspection and post-mortem inspection. However, a smooth slaughter and inspection process is essential. Therefore, one should be able to assess prior to slaughter which pigs are suitable for visual meat inspection only, and which need more profound inspection procedures. This study evaluates the usability of the FCI provided by pig producers and considered the possibility for risk ranking of incoming slaughter batches according to the previous meat inspection data and the current FCI. Eighty-five slaughter batches comprising 8954 fattening pigs were randomly selected at a slaughterhouse that receives animals from across Finland. The mortality rate, the FCI and the meat inspection results for each batch were obtained. The current FCI alone provided insufficient and inaccurate information for risk ranking purposes for meat inspection. The partial condemnation rate for a batch was best predicted by the partial condemnation rate calculated for all the pigs sent for slaughter from the same holding in the previous year (p<0.001) and by prior information on cough declared in the current FCI (p=0.02) statement. Training and information to producers are needed to make the FCI reporting procedures more accurate. Historical meat inspection data on pigs slaughtered from the same holdings and well-chosen symptoms/signs for reporting, should be included in the FCI to facilitate the allocation of pigs for visual inspection. The introduced simple scoring system can be easily used for additional information for directing batches to appropriate meat inspection procedures. To

Beliefs concerning the reliability of nuclear power plant in-service inspections; Uskomuksia ydinvoimalaitoksissa suoritettavien tarkastusten luotettavuudesta

Energy Technology Data Exchange (ETDEWEB)

Kettunen, J. [VTT Automation, Espoo (Finland)

1997-01-01

The aim of the study was to investigate belief systems held by the officials responsible for the planning and supervision of NDT operations within the Finnish nuclear industry. They were asked to express their opinions on (1) the reliability of NDT methods in general, (2) the factors influencing the reliability of in-service inspections, and (3) the degree of reliability of the current inspections operations conducted by means of NDT methods in the Finnish nuclear power plants. Another goal of the study was to assess the adequacy of officials` beliefs (or belief systems). The research data was collected by interviewing representatives from Finnish power companies (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations, and the Finnish Centre for Radiation and Nuclear Safety (STUK). The adequacy of the beliefs expressed was assessed by means of the results obtained from international NDT reliability studies and on the basis of interviewees` own justification. (refs.).

Crack characterization for in-service inspection planning

International Nuclear Information System (INIS)

Waale, J.; Ekstroem, P.

1995-12-01

During in-service inspection by non destructive testing the reliability is highly dependent on how the equipment is adjusted to the specific object and to the anticipated crack feature.The crack feature and morphology vary widely between different cracking mechanisms and between material types in which the cracks appear. The major objective of this study was to characterize a number of morphology parameters for common crack mechanism and structure material combinations. Critical morphology parameters are crack orientation, shape, width, surface roughness and branching. The crack parameters were evaluated from failure analyses reported from the nuclear and non-nuclear industry. In addition, a literature review was carried out on crack parameter reports and on failure analysis reports, which were further evaluated. The evaluated crack parameters were plotted and statistically processed in data groups with respect to crack mechanism and material type. The fatigue crack mechanism were classified as mechanical, thermal or corrosion fatigue and stress corrosion crack mechanism as intergranular, transgranular or inter dendritic stress corrosion cracking. Furthermore, some common weld defects were characterized for comparison. The materials were divided into three broad groups, ferritic low alloy steels, stainless steels and nickel base alloys. The results indicate significant differences between crack parameters when comparing data from different crack mechanism/material type combinations. Typical parameter values and scatter were derived for several combinations where the data was sufficient for statistical significance. 10 refs, 105 figs, 14 tabs

Crack characterization for in-service inspection planning

Energy Technology Data Exchange (ETDEWEB)

Waale, J [SAQ Inspection Ltd, Stockholm (Sweden); Ekstroem, P [ABB Atom AB, Vaesteraas (Sweden)

1995-12-01

During in-service inspection by non destructive testing the reliability is highly dependent on how the equipment is adjusted to the specific object and to the anticipated crack feature.The crack feature and morphology vary widely between different cracking mechanisms and between material types in which the cracks appear. The major objective of this study was to characterize a number of morphology parameters for common crack mechanism and structure material combinations. Critical morphology parameters are crack orientation, shape, width, surface roughness and branching. The crack parameters were evaluated from failure analyses reported from the nuclear and non-nuclear industry. In addition, a literature review was carried out on crack parameter reports and on failure analysis reports, which were further evaluated. The evaluated crack parameters were plotted and statistically processed in data groups with respect to crack mechanism and material type. The fatigue crack mechanism were classified as mechanical, thermal or corrosion fatigue and stress corrosion crack mechanism as intergranular, transgranular or inter dendritic stress corrosion cracking. Furthermore, some common weld defects were characterized for comparison. The materials were divided into three broad groups, ferritic low alloy steels, stainless steels and nickel base alloys. The results indicate significant differences between crack parameters when comparing data from different crack mechanism/material type combinations. Typical parameter values and scatter were derived for several combinations where the data was sufficient for statistical significance. 10 refs, 105 figs, 14 tabs.

Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

International Nuclear Information System (INIS)

Anderson, M.T.; Aldrich, D.A.

1990-09-01

In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

Development of In-Service Inspection system for heat transfer tubes in the primary pressurized water cooler in the HTTR

Energy Technology Data Exchange (ETDEWEB)

Shinozaki, Masayuki; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Wada, Shigeyuki

1999-08-01

The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

A Study of Automation for Examination Analysis of Inservice Inspection for Nuclear Power Plant (I)

International Nuclear Information System (INIS)

Kim, W.

1985-01-01

The developing country, KOREA where does not possess the natural resources for traditional energy such as oil and gas, so. The nuclear energy is the most single reliable source available for closing the energy gap. For these reason, It is inavoidable to construct the nuclear power plant and to develop technology related nuclear energy. The rate of operation in large nuclear power facilities depends upon the performance of work system through design and construction, and also the applied technology. Especially, it is the most important element that safety and reliability in operation of nuclear power plant. In view of this aspects, Nuclear power plant is performed severe examinations during perceives and inservice inspection. This study provide an automation of analysis for volumetric examination which is required to nuclear power plant components. It is composed as follows: I. Introduction II. Inservice Inspection of Nuclear Power Plant * General Requirement. * Principle and Methods of Ultrasonic Test. * Study of Flaw Evaluation and Design of Classifying Formula for Flaws. III. Design of Automation for Flaw Evaluation. IV. An Example V. Conclusion In this theory, It is classifying the flaws, the formula of classifying flaws and the design of automation that is the main important point. As motioned the above, Owing to such as automatic design, more time could be allocated to practical test than that of evaluation of defects, Protecting against subjective bias tester by himself and miscalculation by dint of various process of computation. For the more, adopting this method would be used to more retaining for many test data and comparative evaluating during successive inspection intervals. Inspire of limitation for testing method and required application to test components, it provide useful application to flow evaluation for volumetric examination. Owing to the characteristics of nuclear power plant that is highly skill intensive industry and has huge system, the

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

International Nuclear Information System (INIS)

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

1988-01-01

The evaluation and improvement of NDE Reliability for In-service Inspection (ISI) of Light Water Reactors (NDE Reliability) Program at Pacific Northwest Laboratory (PNL) was established to determine the reliability of current ISI techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this NRC program are to: determine the reliability of ultrasonic ISI performed on commercial light-water reactor (LWR) primary systems; determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to ensure a suitably low failure probability using probabilistic fracture mechanics analysis; evaluate the degree of reliability improvement that could be achieved using improved and advanced NDE technique; and recommend revisions to ASME Code, Section XI, and Regulatory Requirements, based on material properties, service conditions, and NDE uncertainties, that will ensure suitably low failure probabilities. The program consists of three basic tasks: a Piping task, a Pressure Vessel task, and an Evaluation and Improvement in NDE Reliability task. The major efforts were concentrated in the Piping task and the Evaluation and Improvement in NDE Reliability task

Link Between RI-ISI and Inspection Qualification: Relationship between Defect Detection Rate and Margin of Detection

International Nuclear Information System (INIS)

Shepherd, Barrie; Goujon, Sophie; Whittle, John

2007-01-01

Quantitative risk-informed in-service inspection (RI-ISI) requires a quantitative measurement of inspection effectiveness if the risk change associated with an inspection is to be determined. Knowing the probability of detection (POD) as a function of defect depth (through wall dimension) would provide ideal information. However the main in-service inspection method for nuclear plant is ultrasonics, for which defect detection capability depends on a wide variety of parameters besides defect depth, such as defect orientation, roughness, location, shape etc. In recognition of this the European approach to inspection qualification is generally based on some combination of technical justification, and practical trials on a relatively limited number of defects. This inspection qualification process involves demonstrating that defects of concern will generate responses in excess of the specified recording level or noise, depending on the inspection. It is not currently designed to quantify the probability with which defects will be detected. The work described in this report has been performed in order to help address the problem of how the information generated during inspection qualification can be used as an input for RI-ISI. The approach adopted has been to recognise that as the defect response increases above the recording or noise level, the probability of detecting defects is likely to increase. The work therefore involved an investigation of the relationship between POD (strictly speaking defect detection rate) and margin of detection. It involved blind manual and automated ultrasonic trials on artificial defects in test plates designed to generate a range of signal responses. The detection rate for defects which provided signals at a particular level above noise or above a recording level was then measured. A relationship between defect detection rate and margin of detection has been established based on these trials. In addition to establishing a stronger link

Application of non-destructive testing and in-service inspection to research reactors. Results of a co-ordinated research project

International Nuclear Information System (INIS)

2001-12-01

-destructive testing (NDT), are generally called in-service inspections (ISI) and, together with the above specific techniques, are the subject of the present TECDOC. The main objectives of the TECDOC are to present a number of these special techniques and to give guidance for their application. The guidance and recommendations given in this publication form the basis for the conduct of ISI of research reactors with limited hazard potential to the public. This TECDOC is based on the results of a Co-ordinated Research Project (CRP) on the Application of Non-destructive Testing and In-service Inspection to Research Reactors that the IAEA organized in 1995 to supplement its activities on research reactor ageing within its Research Reactor Safety Programme (RRSP). Because of the importance of such in-service inspections within the programmes for the management of ageing in research reactors, this TECDOC will be useful to a large fraction of the currently operating research reactors that are over 30 years old

Application of non-destructive testing and in-service inspections to research reactors and preparation of ISI programme and manual for WWR-C research reactors

International Nuclear Information System (INIS)

Khattab, M.

1996-01-01

The present report gives a review on the results of application of non-destructive testing and in-service inspections to WWR-C reactors in different countries. The major problems related to reactor safety and the procedure of inspection techniques are investigated to collect the experience gained from this type of reactors. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of their rehabilitation programmes. 9 figs., 4 tabs

Information-Driven Inspections

International Nuclear Information System (INIS)

Laughter, Mark D.; Whitaker, J. Michael; Lockwood, Dunbar

2010-01-01

New uranium enrichment capacity is being built worldwide in response to perceived shortfalls in future supply. To meet increasing safeguards responsibilities with limited resources, the nonproliferation community is exploring next-generation concepts to increase the effectiveness and efficiency of safeguards, such as advanced technologies to enable unattended monitoring of nuclear material. These include attribute measurement technologies, data authentication tools, and transmission and security methods. However, there are several conceptual issues with how such data would be used to improve the ability of a safeguards inspectorate such as the International Atomic Energy Agency (IAEA) to reach better safeguards conclusions regarding the activities of a State. The IAEA is pursuing the implementation of information-driven safeguards, whereby all available sources of information are used to make the application of safeguards more effective and efficient. Data from continuous, unattended monitoring systems can be used to optimize on-site inspection scheduling and activities at declared facilities, resulting in fewer, better inspections. Such information-driven inspections are the logical evolution of inspection planning - making use of all available information to enhance scheduled and randomized inspections. Data collection and analysis approaches for unattended monitoring systems can be designed to protect sensitive information while enabling information-driven inspections. A number of such inspections within a predetermined range could reduce inspection frequency while providing an equal or greater level of deterrence against illicit activity, all while meeting operator and technology holder requirements and reducing inspector and operator burden. Three options for using unattended monitoring data to determine an information-driven inspection schedule are to (1) send all unattended monitoring data off-site, which will require advances in data analysis techniques to

In-service inspection for nuclear power plants

International Nuclear Information System (INIS)

1980-01-01

The Safety Guides are recommendations issued by the Agency for use by Member States in the context of their own nuclear safety requirements. Design consideration, in-service examination, test requirements, repair and replacement, equipment, methods and techniques and also administrative aspects are given in this issue

Application of risk-based methods to inservice testing of check valves

Energy Technology Data Exchange (ETDEWEB)

Closky, N.B.; Balkey, K.R.; McAllister, W.J. [and others

1996-12-01

Research efforts have been underway in the American Society of Mechanical Engineers (ASME) and industry to define appropriate methods for the application of risk-based technology in the development of inservice testing (IST) programs for pumps and valves in nuclear steam supply systems. This paper discusses a pilot application of these methods to the inservice testing of check valves in the emergency core cooling system of Georgia Power`s Vogtle nuclear power station. The results of the probabilistic safety assessment (PSA) are used to divide the check valves into risk-significant and less-risk-significant groups. This information is reviewed by a plant expert panel along with the consideration of appropriate deterministic insights to finally categorize the check valves into more safety-significant and less safety-significant component groups. All of the more safety-significant check valves are further evaluated in detail using a failure modes and causes analysis (FMCA) to assist in defining effective IST strategies. A template has been designed to evaluate how effective current and emerging tests for check valves are in detecting failures or in finding significant conditions that are precursors to failure for the likely failure causes. This information is then used to design and evaluate appropriate IST strategies that consider both the test method and frequency. A few of the less safety-significant check valves are also evaluated using this process since differences exist in check valve design, function, and operating conditions. Appropriate test strategies are selected for each check valve that has been evaluated based on safety and cost considerations. Test strategies are inferred from this information for the other check valves based on similar check valve conditions. Sensitivity studies are performed using the PSA model to arrive at an overall IST program that maintains or enhances safety at the lowest achievable cost.

Technical evaluation report on the Third 10-year Interval Inservice Inspection Program Plan: Florida Power and Light Company, Turkey Point Nuclear Power Plant, Units 3 and 4 (Docket Numbers 50-250 and 50-251)

International Nuclear Information System (INIS)

Brown, B.W.; Feige, E.J.; Galbraith, S.G.; Porter, A.M.

1995-02-01

This report presents the results of the evaluation of the Turkey Point Nuclear Power Plant, Units 3 and 4, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted September 9, 1993, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, requirements that the licensee has determined to be impractical. The Turkey Point Nuclear Power Plant, Units 3 and 4, Third 10-Year Interval Inservice Inspection Program Plan is evaluated in Section 2 of this report. The inservice inspection (ISI) program plan is evaluated for (a) compliance with the appropriate edition/addenda of Section XI, (b) acceptability of the examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission (NRC) reviews. The requests for relief are evaluated in Section 3 of this report

Case history of a 94 MVA turbo-generator retired after 190.000 hours of service by defects revealed by boresonic in-service inspection

International Nuclear Information System (INIS)

Porro, F.; Santoro, M.

1990-01-01

The case-history of a turbogenerator manufactured by Ansaldo on 1957 and turned on operation on 1958 then retired after 30 years of operation, with a total of 190.000 hours of service, by defects revealed trough boresonic inspections, will be presented. The rotor was inspected a first time after 130.000 hours of service and was overbored in order to allow further service operations. After other 60.000 hours of service operation the rotor underwent to a new in-service inspection that showed an unacceptable condition. The rotor, retired from service, has been destined to destructive tests in order to verify non-destructive predictions

Risk based inspection experience from the European chemical- and petrochemical industries

International Nuclear Information System (INIS)

Kristensen, Hans; Jeppesen, Leif; Larsen, Bjarne; Kim, Na Yon

2001-01-01

As an inspection vendor with 60 years of experience and with more than 25 years of experience as manufacturer of the Automated Ultrasonic NDT systems, the P-scan systems, FORCE Institute is continuously analysing the market for NDT. This is done to assure that both the equipment product line and the service mix provided by FORCE Institute are meeting the requirements from the industry today and in the future. The concept of Risk Based Inspection Programmes were adopted early by the offshore industry and has in the recent years been adopted by many other industries as a reliable and cost efficient way of maintaining a production facility. A Risk/Reliability Based Inspection Programme is a 'living organism' that constantly needs information if it shall be of any value and NDT information is only one type of information that is required. The NDT information required is normally related to corrosion/base material information and weld integrity information. NDT as an integrated part of a plants maintenance system is, in Europe, currently influenced by the following tendencies which all are related to 'Risk Based Inspection': · Increased use of Base-Line Inspections · Reduction in the use of repeated inspections(qualitative- instead of quantitative Inspections). · Inspection results are fed directly into the plant maintenance system. · Fitness-for-Purpose acceptance criteria instead of conventional acceptance criteria. As repeatability and accuracy is a key issue for the data, automated ultrasonic inspection is increasingly used as an alternative to manual ultrasonic inspection, but due to the physical size of most automated ultrasonic inspection systems the gain in productivity has not been as significant as the gain in repeatability and accuracy. In this paper some of FORCE Institute's practical experiences with examinations carried out in connection with Risk Based Inspection is used to illustrate the above described tendencies. Not only examples using automated

Determining prestressing forces for inspection of prestressed concrete containments

International Nuclear Information System (INIS)

1990-07-01

General Design Criterion 53, ''Provisions for Containment Testing and Inspection,'' of Appendix A, ''General Design Criteria for Nuclear Power Plants,'' to 10 CFR Part 50, ''Domestic Licensing of Production and Utilization Facilities,'' requires, in part, that the reactor containment be designed to permit (1) periodic inspection of all important areas and (2) an appropriate surveillance program. Regulatory Guide 1.35, ''Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures,'' describes a basis acceptable to the NRC staff for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete containment structures of light-water-cooled reactors. This guide expands and clarifies the NRC staff position on determining prestressing forces to be used for inservice inspections of prestressed concrete containment structures

Probabilistic Fatigue Life Updating for Railway Bridges Based on Local Inspection and Repair.

Science.gov (United States)

Lee, Young-Joo; Kim, Robin E; Suh, Wonho; Park, Kiwon

2017-04-24

Railway bridges are exposed to repeated train loads, which may cause fatigue failure. As critical links in a transportation network, railway bridges are expected to survive for a target period of time, but sometimes they fail earlier than expected. To guarantee the target bridge life, bridge maintenance activities such as local inspection and repair should be undertaken properly. However, this is a challenging task because there are various sources of uncertainty associated with aging bridges, train loads, environmental conditions, and maintenance work. Therefore, to perform optimal risk-based maintenance of railway bridges, it is essential to estimate the probabilistic fatigue life of a railway bridge and update the life information based on the results of local inspections and repair. Recently, a system reliability approach was proposed to evaluate the fatigue failure risk of structural systems and update the prior risk information in various inspection scenarios. However, this approach can handle only a constant-amplitude load and has limitations in considering a cyclic load with varying amplitude levels, which is the major loading pattern generated by train traffic. In addition, it is not feasible to update the prior risk information after bridges are repaired. In this research, the system reliability approach is further developed so that it can handle a varying-amplitude load and update the system-level risk of fatigue failure for railway bridges after inspection and repair. The proposed method is applied to a numerical example of an in-service railway bridge, and the effects of inspection and repair on the probabilistic fatigue life are discussed.

Statistical Sampling For In-Service Inspection Of Liquid Waste Tanks At The Savannah River Site

International Nuclear Information System (INIS)

Harris, S.

2011-01-01

Savannah River Remediation, LLC (SRR) is implementing a statistical sampling strategy for In-Service Inspection (ISI) of Liquid Waste (LW) Tanks at the United States Department of Energy's Savannah River Site (SRS) in Aiken, South Carolina. As a component of SRS's corrosion control program, the ISI program assesses tank wall structural integrity through the use of ultrasonic testing (UT). The statistical strategy for ISI is based on the random sampling of a number of vertically oriented unit areas, called strips, within each tank. The number of strips to inspect was determined so as to attain, over time, a high probability of observing at least one of the worst 5% in terms of pitting and corrosion across all tanks. The probability estimation to determine the number of strips to inspect was performed using the hypergeometric distribution. Statistical tolerance limits for pit depth and corrosion rates were calculated by fitting the lognormal distribution to the data. In addition to the strip sampling strategy, a single strip within each tank was identified to serve as the baseline for a longitudinal assessment of the tank safe operational life. The statistical sampling strategy enables the ISI program to develop individual profiles of LW tank wall structural integrity that collectively provide a high confidence in their safety and integrity over operational lifetimes.

In-service Inspection of Radioactive Waste Tanks at the Savannah River Site – 15410

Energy Technology Data Exchange (ETDEWEB)

Wiersma, Bruce [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maryak, Matthew [Savannah River Remediation, LLC., Aiken, SC (United States); Baxter, Lindsay [Univ. of Notre Dame, IN (United States); Harris, Stephen [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Elder, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-01-12

Liquid radioactive wastes from the Savannah River Site (SRS) separation process are stored in large underground carbon steel tanks. The high level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks will be required to be in service for times exceeding the initial intended life. Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. A structural integrity program is in place to maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. In-service inspection (ISI) is an essential element of a comprehensive structural integrity program for the waste tanks at the Savannah River Site (SRS). The ISI program was developed to determine the degree of degradation the waste tanks have experienced due to service conditions. As a result of the inspections, an assessment can be made of the effectiveness of corrosion controls for the waste chemistry, which precludes accelerated localized and general corrosion of the waste tanks. Ultrasonic inspections (UT) are performed to detect and quantify the degree of general wall thinning, pitting and cracking as a measure of tank degradation. The results from these inspections through 2013, for the 27 Type III/IIIA tanks, indicate no reportable in-service corrosion degradation in the primary tank (i.e., general, pitting, or cracking). The average wall thickness for all tanks remains above the manufactured nominal thickness minus 0.25 millimeter and the largest pit identified is

Automatic ultrasonic pre-service, and in-service inspection of pressurized components of the primary circuit of nuclear power stations

International Nuclear Information System (INIS)

Muller, G.P.; Hallermeier, L.; Heinrich, D.; Grabendorfer, W.; Rebrmann, M.

1985-01-01

Ultrasonic pre-service and especially in-service inspection activities on the primary circuit of nuclear power stations form an essential part of the maintenance work that must be performed throughout the lifetime to ensure plant integrity. Consequently, the equipment required to carry out these inspections must be continuously improved in respect of reliability, safety, accuracy and ease of handling in order to minimize disturbances and repairs and reduce radiation exposure of the personnel. The authors' discussion of technique, equipment and performance of automated ultrasonic inspection is based on 15 years of experience in the testing of components of the primary circuit in nuclear power stations. To cover all inspection areas of the RPV of a PWR, four different manipulators are required, two for the closure head, one for the studs and one for the cylindrical shell and bottom closure. The use of the newly developed equipment, which naturally meets all the recommendations of the licensing authorities, allows for the automatic inspection of the components of primary circuit of nuclear power stations and the thus helps to substantially decrease the radiation exposure of the personnel. All the manipulators and their control consoles were designed and manufactured by M.A.N., Nuremberg while the ultrasonic electronic system was developed by Krautkramer, Cologne

Integration of in-service inspection works in the objectives of nuclear power plants; Integracion de los trabajos de inspeccion en servicio en los objectivos de las centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Garcia, M.; Lopez, M.A.

1996-09-01

The present articles summarizes the objectives of Spanish NPPs and the in-service inspection aspect in these objectives. The Safety maintenance, lifetime of nuclear power plants, reduction of doses and wastes and the participation of main are evaluated.

In-service inspection of electronics components, circuits and nuclear radiation detectors

International Nuclear Information System (INIS)

Darbhe, M.D.

2002-01-01

A nuclear reactor is a complex process plant. Like a nuclear power plant, the research reactors also employ various nuclear and process systems, the scope and number of such systems being plant-specific. In-service inspection of these systems is an important requirement and is applied at various levels of their constituent units such as detectors, electronics components, circuits and integrated systems. The sensors used cover a wide range such as neutronic, radiation, process (pressure, temperature, flow, level) and many others. The present discussion is limited to neutronic and radiation detectors. The electronic components used normally consist of passive components like resistors, capacitors, semiconductor components like diodes, transistors, analog integrated circuits and digital integrated circuits and electromagnetic relays, to name a few. In order to have a comprehensive surveillance and ISI plan, over the entire plant life, it is necessary to understand various mechanisms, which degrade the performance of these systems. These are discussed initially and later various ISI methods that are used on component-circuit or system level, to ensure optimum system performance, are discussed. The computerised systems, because of hardware and software considerations, have to be given special attention, and the same are discussed briefly

Risk-based inspection and maintenance systems for steam turbines

International Nuclear Information System (INIS)

Fujiyama, Kazunari; Nagai, Satoshi; Akikuni, Yasunari; Fujiwara, Toshihiro; Furuya, Kenichiro; Matsumoto, Shigeru; Takagi, Kentaro; Kawabata, Taro

2004-01-01

The risk-based maintenance (RBM) system has been developed for steam turbine plants coupled with the quick inspection systems. The RBM system utilizes the field failure and inspection database accumulated over 30 years. The failure modes are determined for each component of steam turbines and the failure scenarios are described as event trees. The probability of failure is expressed in the form of unreliability functions of operation hours or start-up cycles through the cumulative hazard function method. The posterior unreliability is derived from the field data analysis according to the inspection information. Quick inspection can be conducted using air-cooled borescope and heat resistant ultrasonic sensors even if the turbine is not cooled down sufficiently. Another inspection information comes from degradation and damage measurement. The probabilistic life assessment using structural analysis and statistical material properties, the latter is estimated from hardness measurement, replica observation and embrittlement measurement. The risk function is calculated as the sum product of unreliability functions and expected monetary loss as the consequence of failure along event trees. The optimum maintenance plan is determined among simulated scenarios described through component breakdown trees, life cycle event trees and risk functions. Those methods are effective for total condition assessment and economical maintenance for operating plants

ASME N511-19XX, Standard for periodic in-service testing of nuclear air treatment, heating, ventilating and air conditioning systems

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-08-01

A draft version of the Standard is presented in this document. The Standard covers the requirements for periodic in-service testing of nuclear safety-related air treatment, heating, ventilating, and air conditioning systems in nuclear facilities. The Standard provides a basis for the development of test programs and does not include acceptance criteria, except in cases where the results of one test influence the performance of other tests. The Standard covers general inspection and test requirements, reference values, inspection and test requirements, generic tests, acceptance criteria, in-service test requirements, testing following an abnormal incident, corrective action requirements, and quality assurance. Mandatory appendices provide a visual inspection checklist and four test procedures. Non-mandatory appendices provide additional information and guidance on mounting frame pressure leak test procedure, corrective action, challenge gas substitute selection criteria, and test program development. 8 refs., 10 tabs.

ASME N511-19XX, Standard for periodic in-service testing of nuclear air treatment, heating, ventilating and air conditioning systems

International Nuclear Information System (INIS)

1997-01-01

A draft version of the Standard is presented in this document. The Standard covers the requirements for periodic in-service testing of nuclear safety-related air treatment, heating, ventilating, and air conditioning systems in nuclear facilities. The Standard provides a basis for the development of test programs and does not include acceptance criteria, except in cases where the results of one test influence the performance of other tests. The Standard covers general inspection and test requirements, reference values, inspection and test requirements, generic tests, acceptance criteria, in-service test requirements, testing following an abnormal incident, corrective action requirements, and quality assurance. Mandatory appendices provide a visual inspection checklist and four test procedures. Non-mandatory appendices provide additional information and guidance on mounting frame pressure leak test procedure, corrective action, challenge gas substitute selection criteria, and test program development. 8 refs., 10 tabs

Inspection qualification and implementation of ENIQ in Sweden

International Nuclear Information System (INIS)

Zettervall, Tommy

2013-01-01

Many countries are currently considering their approaches to inspection qualification and risk-informed in-service inspection (RI-ISI) and are carefully assessing experience data. In Europe most of the utilities operating nuclear power plants have joined together to form the European Network for Inspection Qualification ENIQ. In practice, qualification can be performed with varying degrees of complexity and cost, varying from capability statement based on existing evidence, through to an extensive qualification consisting of a detailed Technical Justification (TJ) together with open and blind trials on full-scale test blocks. An Inspection Qualification is an investigation and demonstration, which confirm that an inspection system has the ability to solve its specific tasks. The qualification is a Quality Assurance of an inspection system based on documents and practical trials. A reliable inspection system, based on a reliable qualification and correct prerequisites, will reduce total costs for Licensees and increase the credibility of the inspection result. To get such inspection system, which could be valid for many years, it's of necessity to fulfil all included parts in the process. It begins with the Technical Specification from Licensee, where input data and requirements about the actual component are specified. To get an inspection system that could live over time, the Technical Justification is of importance. Finally the test blocks and used simulation techniques play an important part of the final result, and these test blocks together with TJ form the basis for qualification body's decision. (author)

A study on the effect of flaw detection probability assumptions on risk reduction achieved by non-destructive inspection

International Nuclear Information System (INIS)

Cronvall, O.; Simola, K.; Männistö, I.; Gunnars, J.; Alverlind, L.; Dillström, P.; Gandossi, L.

2012-01-01

Leakages and ruptures of piping components lead to reduction or loss of the pressure retaining capability of the system, and thus contribute to the overall risk associated with nuclear power plants. In-service inspection (ISI) aims at verifying that defects are not present in components of the pressure boundary or, if defects are present, ensuring that these are detected before they affect the safe operation of the plant. Reliability estimates of piping are needed e.g., in probabilistic safety assessment (PSA) studies, risk-informed ISI (RI-ISI) applications, and other structural reliability assessments. Probabilistic fracture mechanics models can account for ISI reliability, but a quantitative estimate for the latter is needed. This is normally expressed in terms of probability of detection (POD) curves, which correlate the probability of detecting a flaw with flaw size. A detailed POD curve is often difficult (or practically impossible) to obtain. If sufficient risk reduction can be shown by using simplified (but reasonably conservative) POD estimates, more complex PODs are not needed. This paper summarises the results of a study on the effect of piping inspection reliability assumptions on failure probability using structural reliability models. The main interest was to investigate whether it is justifiable to use a simplified POD curve. Further, the study compared various structural reliability calculation approaches for a set of analysis cases. The results indicate that the use of a simplified POD could be justifiable in RI-ISI applications.

Qualification of NDT systems for in-service inspections of nuclear power plant pressure vessels; Ydinvoimalaitosten paineastioiden maeaeraeaikaistarkastuksissa kaeytettaevien NDT-jaerjestelmien paetevoeinti

Energy Technology Data Exchange (ETDEWEB)

Elfving, K

1998-11-01

The goal of this study is to determine the requirements of the in-service inspection qualification in Europe, their feasibility in practice and to find out possible manufacture defects in test pieces used in practical trials. The literature study consists of qualification requirements set by European regulatory bodies and by the European nuclear power utilities. Also a brief summary of qualification requirements set by ASME Code, Section XI and comparison between ASME and European qualification requirements is included 24 refs.

Examination of in-service risk management at the level of the draft project for Cigeo

International Nuclear Information System (INIS)

2014-01-01

This document reports the opinion of the IRSN on the issue of in-service risk management as it appears in the draft project of Cigeo. After having recalled the general context of the ASN request for this opinion, and presented the ANDRA file on in-service risk-management of the draft project for Cigeo, this report describes the underground installation and the different flows during the exploitation phase, and recalls the various evolutions of Cigeo general design from the 2009 file to the draft project. The next part reports the examination by the IRSN of the relevance of the in-service safety approach by the ANDRA. The next chapter reports the examination of the risk analysis made by the ANDRA on radioactive material dissemination, fire (protection measures, ventilation control, case of exothermal reactions), explosion, handling, co-activity, as well as ageing and maintenance of storage components, intervention in accidental and incidental situation and possibility of retrieval of waste packages

Reproducibility problems of in-service ultrasonic testing results

International Nuclear Information System (INIS)

Honcu, E.

1974-01-01

The reproducibility of the results of ultrasonic testing is the basic precondition for its successful application in in-service inspection of changes in the quality of components of nuclear power installations. The results of periodic ultrasonic inspections are not satisfactory from the point of view of reproducibility. Regardless, the ultrasonic pulse-type method is suitable for evaluating the quality of most components of nuclear installations and often the sole method which may be recommended for inspection with regard to its technical and economic aspects. (J.B.)

Automation of eddy current system for in-service inspection of turbine and generator rotor bores

International Nuclear Information System (INIS)

Anon.

1987-01-01

The current project (EPRI-RP-1975-5) is a continuation of activities that began several years ago. Those results (EPRI-RP-1957-1) indicated that eddy current testing shows promise for in-service inspection. The current project investigates the degree to which eddy current testing can be used to replace bore magnetic particle testing. For this purpose, correlation studies between eddy current and magnetic particle tests are being undertaken on laboratory rotor sections and field test pieces of rotors. The eddy current data are to be gathered automatically by a combination of the Nortec-25L Eddyscope (to provide the analog eddy current signals) and the General Electric DATAQ/sup TM/ (a registered trademark of the General Electric Co.) System (to perform the automatic data acquisition). This paper describes some test results on a flaked laboratory rotor section

Fuel channel in-service inspection programs program design for maximum cost effectiveness

International Nuclear Information System (INIS)

Van den Brekel, N.C.

1995-01-01

Inspection is an integral part of fuel channel life management strategy. Inspection data is used to assess the state of reactor core integrity and provide the information necessary to optimize long term maintenance programs. This paper will provide an overview of the structured approach to developing fuel channel inspection programs within OHN. The inspection programs are designed to balance the resources utilized (cost, outage time, and dose expenditure) with the benefits provided by the inspection data obtained (improved knowledge of component status, degradation mechanisms and rates, etc..). The CANDU community has yet to have a fuel channel operate for a full 30 year design life. Since research programs can not fully simulate reactor operating conditions, inspections become an essential feature of the life management strategy as the components age. Inspection programs often include activities designed to develop predictive capability for long term fuel channel behaviour and provide early warning of changes in behaviour. It should be noted that although this paper addresses the design of fuel channel inspection programs, the basic principles presented can be applied to the design of inspection programs for any major power plant component or system. (author)

PRA and Risk Informed Analysis

International Nuclear Information System (INIS)

Bernsen, Sidney A.; Simonen, Fredric A.; Balkey, Kenneth R.

2006-01-01

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. The approach allows calculations of conditional core damage probabilities both for component failures that cause initiating events and failures in standby systems that decrease the availability of these systems. The paper relates the explicit risk based methods of the new Section XI code cases to the implicit consideration of risk used in the development of Section XI. Other topics include the needed interactions of ISI engineers, plant operating staff, PRA specialists, and members of expert panels that review the risk based programs

The broken food chain information: cross border issues of risk-based meat inspection in the pig sector in Germany and the Netherlands

NARCIS (Netherlands)

Wilke, T.; Wagenberg, van C.P.A.; Oosterkamp, E.B.; Bondt, N.; Lang, J.

2012-01-01

European food safety legislation allows for a new type of meat inspection in swine. This ‘risk-based meat inspection without incision’ or ‘supply chain meat inspection’ (SCMI) uses food chain information to derive a veterinary prognosis on the pigs’ health prior to delivery to the slaughterhouse and

Automated ultrasonic inspection of nuclear plant components

International Nuclear Information System (INIS)

Baron, J.A.; Dolbey, M.P.

1982-01-01

For reasons of safety and efficiency, automated systems are used in performing ultrasonic inspection of nuclear components. An automated system designed specifically for the inspection of headers in a nuclear plant is described. In-service inspection results obtained with this system are shown to correlate with pre-service inspection results obtained by manual methods

Current activities and challenges of the European network for inspection and qualification (ENIQ)

Energy Technology Data Exchange (ETDEWEB)

Martin, Oliver [European Commission, Joint Research Centre (JRC), Petten (Netherlands). Institute for Energy and Transport (IET); Martin, Etienne [EDF, St Denis (France). Direction Production Ingenierie; Booler, Russ [AMEC Clean Energy Europe, Warrington (United Kingdom); Zetterwall, Tommy [Swedish Qualification Centre, Taeby (Sweden); Walker, Tony [Rolls-Royce Submarines, Derby (United Kingdom)

2014-10-15

This article describes the development of the European Network for Inspection and Qualification (ENIQ) since the previous presentation of the network in the 2011 April/May edition of this journal, covering mainly the new technical challenges facing the network and resulting projects as well as the establishment of the new Task Group for Inspection Qualification Bodies. ENIQ is a utility-driven network dealing with the reliability and effectiveness of non-destructive testing (NDT) for nuclear power plants (NPP). ENIQ is recognised as one of the main contributors to today's global qualification codes and guidelines for in-service inspection (ISI) and has published nearly 50 documents. Among them are the 'European Methodology for Qualification of Non-Destructive Testing', the first qualification methodology based on technical justifications, the 'European Framework Document for Risk-Informed In-Service Inspection (RI-ISI)', and various recommend practices. In addition ENIQ has carried out two pilot studies and a number of surveys. In 2012, ENIQ joined the European based R and D association on Gen II and III reactors, NUGENIA, making ENIQ its 8{sup th} technical area. Following the entry into NUGENIA, ENIQ members have updated the ENIQ roadmap and included a number of new technical challenges facing its members in the near future. Also ENIQ established a third task group in 2013, the Task Group for Inspection Qualification Bodies (TGIQB), which should serve as an exchange forum for inspection qualification bodies. ENIQ is currently preparing or performing new projects and studies to tackle these challenges and new recommended practices and reports are likely to evolve from these projects, which will enable ENIQ to maintain its role as one of the main contributors to today's global qualification codes and guidelines for ISI.

Current activities and challenges of the European network for inspection and qualification (ENIQ)

International Nuclear Information System (INIS)

Martin, Oliver; Martin, Etienne; Zetterwall, Tommy; Walker, Tony

2014-01-01

This article describes the development of the European Network for Inspection and Qualification (ENIQ) since the previous presentation of the network in the 2011 April/May edition of this journal, covering mainly the new technical challenges facing the network and resulting projects as well as the establishment of the new Task Group for Inspection Qualification Bodies. ENIQ is a utility-driven network dealing with the reliability and effectiveness of non-destructive testing (NDT) for nuclear power plants (NPP). ENIQ is recognised as one of the main contributors to today's global qualification codes and guidelines for in-service inspection (ISI) and has published nearly 50 documents. Among them are the 'European Methodology for Qualification of Non-Destructive Testing', the first qualification methodology based on technical justifications, the 'European Framework Document for Risk-Informed In-Service Inspection (RI-ISI)', and various recommend practices. In addition ENIQ has carried out two pilot studies and a number of surveys. In 2012, ENIQ joined the European based R and D association on Gen II and III reactors, NUGENIA, making ENIQ its 8 th technical area. Following the entry into NUGENIA, ENIQ members have updated the ENIQ roadmap and included a number of new technical challenges facing its members in the near future. Also ENIQ established a third task group in 2013, the Task Group for Inspection Qualification Bodies (TGIQB), which should serve as an exchange forum for inspection qualification bodies. ENIQ is currently preparing or performing new projects and studies to tackle these challenges and new recommended practices and reports are likely to evolve from these projects, which will enable ENIQ to maintain its role as one of the main contributors to today's global qualification codes and guidelines for ISI.

In-service materials testing of selected components of unit 1 and 2 of V-1 nuclear power plant

International Nuclear Information System (INIS)

Cintula, J.

1982-01-01

The task of in-service nondestructive testing of nuclear installations is to confirm that the state of base material and welded joints has not changed owing to mechanical, thermal or radiation stress. Under the regulations of safe operation the first in-service inspection of all components of a WWER 440 reactor must be carried out after 15,000 to 2O,00O operating hours at the latest. Further in-service inspections are repeated after 30,000 hours (pressure vessels) and 40,000 hours (the main steam piping and the feedwater piping). Proceeding from experience gained so far, intervals are suggested for in-service checks of the other components of the V-1 nuclear power plant. Also briefly described are the main nondestructive methods used for such checks at this power plant. (Z.M.)

Risk-based methodology for USNRC inspections

International Nuclear Information System (INIS)

Wong, S.M.; Holahan, G.M.; Chung, J.W.; Johnson, M.R.

1995-01-01

This paper describes the development and trial applications of a risk-based methodology to enhance the inspection processes for US nuclear power plants. Objectives of risk-based methods to complement prescriptive engineering approaches in US Nuclear Regulatory Commission (USNRC) inspection programs are presented. Insights from time-dependent risk profiles of plant configurational from Individual Plant Evaluation (IPE) studies were integrated to develop a framework for optimizing inspection efforts in NRC regulatory initiatives. Lessons learned from NRC pilot applications of the risk-based methodology for evaluation of the effectiveness of operational risk management programs at US nuclear power plant sites are also discussed

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

Risk-based inspection of nuclear power plants

International Nuclear Information System (INIS)

Masopust, R.

1995-01-01

A multidiscipline research programme was developed in the USA to establish risk-based inspections for NPP structures and equipment components. Based on this US research effort, the risk-based procedure for developing inspection guidelines for NPPs is described. The procedure includes the definition of systems, qualitative risk assessment, qualitative risk analysis and development of the inspection programme. The method, when adopted and modified, is recommended also for risk-based inspections of structures and equipment of WWER-type NPPs. A pilot application of the method to unit 1 of the Surry NPP is summarized. (Z.S.) 1 tab., 1 fig., 5 refs

Demonstration test of in-service inspection methods

International Nuclear Information System (INIS)

Takumi, Kenji

1987-01-01

The major objectives of the project are: (1) to demonstrate the reliability of a manual ultrasonic flaw detector and techniques that are used in operating light water reactor plants and (2) to demonstrate the performance and reliability of an automatic ultrasonic flaw detector that is designed to shorten the time required for ISI work and reduce the exposure risk of inspection personnel. The test project consists of three stages. In the first stage, which ended in 1982, defects were added intentionally to a model structure the same in size as a typical 1.1 million kW BWR plant and manual ultrasonic flaw detection tensting was performed. In the second stage, completed in 1984, automatic eddy-current flaw detection testing was carried out for defects in heat transfer piping of a PWP steam generator. In the third stage, which started in 1981 and ended in March 1987, a newly developed automatic ultrasonic flaw detector is applied to testing of defects used for the manual detector performance evaluation. Results have shown that the automatic eddy-current flaw detector under test has an adequately stable performance for practical uses, with a very high reproducibility to permit close inspection of secular deterioration in heat transfer pipes. It has also revealed that both the manual and automatic ultrasonic flaw detectors under test can detect all defects that do not comply with the ASME standards. (Nogami, K.)

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

Development and application of proposed ASME Section XI Code changes for risk-based inspection of piping

International Nuclear Information System (INIS)

West, R.A.

1996-01-01

This synopsis has been written to describe a perspective on the development and application of ASME Section XI Code changes for risk-based inspection of piping. The content is specifically related to the use of risk-based technology for Inservice Inspection (ISI) of piping and efforts made to support the ASME Research/Westinghouse Owners Group/Millstone Unit 3 approach for use of this technology. The opinions contained herein may or may not reflect those of the ASME Codes and Standards Committees responsible for these activities. In order to take such a detailed technical subject and put it into an understandable format, the author has chosen to provide an analogy to simplify what is actually taking place. Risk-based technology in the ISI of piping can be likened to the process of making and using specifically ground prescription glasses to allow for better vision. It provides a process to develop and use these uniquely ground glasses that will dynamically focus on all the locations and obstacles within a plant's piping systems that could cause that plant to trip and fall; more importantly it identifies the locations where the fall could possibly hurt someone else. In this way, Nuclear Safety is being addressed

A web environment for inspection service

International Nuclear Information System (INIS)

Robles Guemes, N.; Garcia Heras, M.; Cueto-Felgueroso Garcia, C.; Pelaez Gutierrez, J. A.

2013-01-01

The in-service inspection fundamental objective is to maintain the structural integrity of structures, systems and components during the life in the installation service. Nuclear power plants are required to implement a Service inspection program in accordance with the established regulation by the Regulatory body.

Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

International Nuclear Information System (INIS)

Singh, Ashutosh Pratap; Rajagopalan, C.; Rakesh, V.; Rajendran, S.; Venugopal, S.; Kasiviswanathan, K.V.; Jayakumar, T.

2011-01-01

Highlights: → Conceptual study on the configuration of an ISI device for FBR interspace environment has been carried out. → Prototyping of the concept has been experimentally validated in a mock up. → High temperature version of the ISI device has been made and tested in mock-up. Further experimentation is underway. → Simulation of different configurations of the device has been carried out with respect to reduced gap between main vessel and safety vessel for future FBRs. → Studies on wheel lining for the device have been carried out at 150 o C for better traction and payload capability. - Abstract: In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.

Designing for in-service inspection

International Nuclear Information System (INIS)

Dixit, K.B.

2002-01-01

The process of design of components, equipment and systems normally focuses on functional requirements, integrity and environmental conditions. Manufacturing processes, fabricability, inspectability and maintainability in service are considered while designing. Provisions for the requirements of ISI also need to be given due consideration while designing the systems, equipment and components

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.)

Human performance in nondestructive inspections and functional tests: Final report

International Nuclear Information System (INIS)

Harris, D.H.

1988-10-01

Human performance plays a vital role in the inspections and tests conducted to assure the physical integrity of nuclear power plants. Even when technically-sophisticated equipment is employed, the outcome is highly dependent on human control actions, calibrations, observations, analyses, and interpretations. The principal consequences of inadequate performance are missed or falsely-reported defects. However, the cost-avoidance that stems from addressing potential risks promptly, and the increasing costs likely with aging plants, emphasize that timeliness and efficiency are important inspection-performance considerations also. Human performance issues were studied in a sample of inspections and tests regularly conducted in nuclear power plants. These tasks, selected by an industry advisory panel, were: eddy-current inspection of steam-generator tubes; ultrasonic inspection of pipe welds; inservice testing of pumps and valves; and functional testing of shock suppressors. Information was obtained for the study from industry and plant procedural documents; training materials; research reports and related documents; interviews with training specialists, inspectors, supervisory personnel, and equipment designers; and first-hand observations of task performance. Eleven recommendations are developed for improving human performance on nondestructive inspections and functional tests. Two recommendations were for the more-effective application of existing knowledge; nine recommendations were for research projects that should be undertaken to assure continuing improvements in human performance on these tasks. 25 refs., 9 figs., 1 tab

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)

Preliminary conceptual design of inspection and maintenance for KALIMER reactor system

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Seok Hun; Yoo, Bong

2000-08-01

In-service inspection and maintenance are very important for improving the safety and availability of nuclear power plants. The conceptual requirements of in-service inspection and maintenance should be reflected in the earlier design process for the verification of the plant operability and reliability. In this report the fundamental approaches of the inspection and maintenance for KALIMER are established to ensure the structural integrity and operability for KALIMER. The general strategy and methodology of maintenance and inspection for the reactor system and components are proposed and described for satisfying the intents of the section XI, division 3, of ASME code and considering the design characteristics of KALIMER.

Present activity in ASME Section XI regarding risk-informed maintenance

International Nuclear Information System (INIS)

Hedden, Owen; Chockie, Alan

2005-01-01

Since 1996 Section XI of the ASME Boiler and Pressure Vessel Code has actively incorporated risk-informed concepts. The risk-informed process provides a framework for allocating inspection resources in a cost-effective manner and helps focus inspections where most critical for plant safety. Based on the success of the risk-informed ISI piping applications at US and non-US plants, Section XI has refined existing Code Cases and expanded the use of the risk-informed process to a variety of high-risk components and systems. The risk informed approach started in the area of inspection and is now being expanded to other plant maintenance activities. This article summarizes the Section XI actions and the continued development of the risk-informed process to improve nuclear plant maintenance. (author)

GASVOL 18'' gas pipeline - risk based inspection study

Energy Technology Data Exchange (ETDEWEB)

Bjoernoey, Ola H.; Etterdal, Birger A. [Det Norske Veritas (DNV), Oslo (Norway); Guarize, Rosimar; Oliveira, Luiz F.S. [Det Norske Veritas (DNV) (Brazil); Faertes, Denise; Dias, Ricardo [TRANSPETRO - PETROBRAS Transporte S.A., Rio de Janeiro, RJ (Brazil)

2003-07-01

This paper describes a risk based approach and inspection planning as part of the Pipeline Integrity Management (PIM) system for the 95.5 km long 18'' GASVOL gas pipeline in the South eastern region of Brazil transporting circa 5 000 000 m3 dry gas per day. Pipeline systems can be subject to several degradation mechanisms and inspection and monitoring are used to ensure system integrity. Modern pipeline regulations and codes are normally based on a core safety or risk philosophy. The detailed design requirements presented in design codes are practical interpretations established so as to fulfill these core objectives. A given pipeline, designed, constructed and installed according to a pipeline code is therefore the realization of a structure, which, along its whole length, meets the applicable safety objectives of that code. The main objective of Pipeline Integrity Management (PIM) is to control and document the integrity of the pipeline for its whole service life, and to do this in a cost-effective manner. DNV has a specific approach to RBI planning, starting with an initial qualitative assessment where pipelines and damage type are ranked according to risk and potential risk reduction by an inspection and then carried forward to a quantitative detailed assessment where the level of complexity and accuracy can vary based on availability of information and owner needs. Detailed assessment requires significant effort in data gathering. The findings are dependent upon the accuracy of the inspection data, and on DNV's interpretation of the pipeline reference system and simplifications in the inspection data reported. The following specific failure mechanisms were investigated: internal corrosion, external corrosion, third party interference, landslides and black powder. RBI planning, in general words, is a 'living process'. In order to optimize future inspections, it is essential that the analyses utilize the most recent information regarding

Value of information-based inspection planning for offshore structures

DEFF Research Database (Denmark)

Irman, Arifian Agusta; Thöns, Sebastian; Leira, Bernt J.

2017-01-01

with each inspection strategy. A simplified and generic risk-based inspection planning utilizing pre- posterior Bayesian decision analysis had been proposed by Faber et al. [1] and Straub [2]. This paper provides considerations on the theoretical background and a Value of Information analysis......-based inspection planning. The paper will start out with a review of the state-of-art RBI planning procedure based on Bayesian decision theory and its application in offshore structure integrity management. An example of the Value of Information approach is illustrated and it is pointed to further research......Asset integrity and management is an important part of the oil and gas industry especially for existing offshore structures. With declining oil price, the production rate is an important factor to be maintained that makes integrity of the structures one of the main concerns. Reliability based...

Ultrasonic inspection reliability for intergranular stress corrosion cracks

Energy Technology Data Exchange (ETDEWEB)

Heasler, P G; Taylor, T T; Spanner, J C; Doctor, S R; Deffenbaugh, J D [Pacific Northwest Lab., Richland, WA (USA)

1990-07-01

A pipe inspection round robin entitled Mini-Round Robin'' was conducted at Pacific Northwest Laboratory from May 1985 through October 1985. The research was sponsored by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research under a program entitled Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors.'' The Mini-Round Robin (MRR) measured the intergranular stress corrosion (GSC) crack detection and sizing capabilities of inservice inspection (ISI) inspectors that had passed the requirements of IEB 83-02 and the Electric Power Research Institute (EPRI) sizing training course. The MRR data base was compared with an earlier Pipe Inspection Round Robin (PIRR) that had measured the performance of inservice inspection prior to 1982. Comparison of the MRR and PIRR data bases indicates no significant change in the inspection capability for detecting IGSCC. Also, when comparing detection of long and short cracks, no difference in detection capability was measured. An improvement in the ability to differentiate between shallow and deeper IGSCC was found when the MRR sizing capability was compared with an earlier sizing round robin conducted by the EPRI. In addition to the pipe inspection round robin, a human factors study was conducted in conjunction with the Mini-Round Robin. The most important result of the human factors study is that the Relative Operating Characteristics (ROC) curves provide a better methodology for describing inspector performance than only probability of detection (POD) or single-point crack/no crack data. 6 refs., 55 figs., 18 tabs.

In service inspection of SUPERPHENIX 1 vessels: MIR

International Nuclear Information System (INIS)

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

1985-01-01

Although no in-service inspection constraints were imposed on the Phenix vessels, the Safety Authorities asked that the design of SUPERPHENIX 1 makes it possible to monitor throughout the lifetime of the reactor, surface and internal defects on the main vessel. A pool design and the presence of heat baffles inside the main vessel make access from the inside of the vessel impossible. Thus, an inspection can only be performed from the outside of the main vessel: the distance between the walls of the main and safety vessels is such that an inspection device can be introduced into the corresponding space. As the design of the reactor precludes radiographic inspection, the method which was selected for monitoring internal defects in the main vessel is ultrasonics. However, the anisotropic structure of austenitic stainless steel welds limits the performance of this technique. The authors present the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of SUPERPHENIX 1 vessels

Inservice inspection of heavy water plants - a tool in assessing damage to components and life extension

International Nuclear Information System (INIS)

Subramanian, C.V.; Thavasimuthu, M.; Bhattacharys, D.K.; Baldev Raj

1994-01-01

Any system and its components are expected to give trouble free service over a certain period of time known as life time. The life time is estimated during the design stage. To achieve the design life, certain level of quality are to be defined and this quality has to be worked into the components by proper fabrication processes and their compliance with quality are to be checked. In addition, one has to guard against initiation or propagation of defects which may occur due to normal and abnormal service conditions. Non-destructive test (NDT) techniques are widely used for finding the health of the component. The role of NDT extends from the production stage to the entire life period of the system. This paper highlights the periodic in-service inspection (ISI) carried out on various components of the Heavy Water Plants (HWP) in India in assessing the integrity of the components and predicting the life of the components. (author). 3 refs., 4 figs

Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

International Nuclear Information System (INIS)

Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E.

1991-09-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab

Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

International Nuclear Information System (INIS)

Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1992-10-01

In a study sponsored by the US Nuclear regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. North Anna was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the North Anna plant

Towards risk-based structural integrity methods for PWRs

International Nuclear Information System (INIS)

Chapman, O.J.V.; Lloyd, R.B.

1992-01-01

This paper describes the development of risk-based structural integrity assurance methods and their application to Pressurized Water Reactor (PWR) plant. In-service inspection is introduced as a way of reducing the failure probability of high risk sites and the latter are identified using reliability analysis; the extent and interval of inspection can also be optimized. The methodology is illustrated by reference to the aspect of reliability of weldments in PWR systems. (author)

Risk-based inservice testing program modifications at Palo Verde nuclear generating station

International Nuclear Information System (INIS)

Knauf, S.; Lindenlaub, B.; Linthicum, R.

1996-01-01

Arizona Public Service Company (APS) is investigating changes to the Palo Verde Inservice Testing (IST) Program that are intended to result in the reduction of the required test frequency for various valves in the American Society of Mechanical Engineers (ASME) Section XI IST program. The analytical techniques employed to select candidate valves and to demonstrate that these frequency reductions are acceptable are risk based. The results of the Palo Verde probabilistic risk assessment (PRA), updated in June 1994, and the risk significant determination performed as part of the implementation efforts for 10 CFR 50.65 (the maintenance rule) were used to select candidate valves for extended test intervals. Additional component level evaluations were conducted by an 'expert panel.' The decision to pursue these changes was facilitated by the ASME Risk-Based Inservice Testing Research Task Force for which Palo Verde is participating as a pilot plant. The NRC's increasing acceptance of cost beneficial licensing actions and risk-based submittals also provided incentive to seek these changes. Arizona Public Service is pursuing the risk-based IST program modification in order to reduce the unnecessary regulatory burden of the IST program through qualitative and quantitative analysis consistent with maintaining a high level of plant safety. The objectives of this project at Palo Verde are as follows: (1) Apply risk-based technologies to IST components to determine their risk significance (i.e., high or low). (2) Apply a combination of deterministic and risk-based methods to determine appropriate testing requirements for IST components including improvement of testing methods and frequency intervals for high-risk significant components. (3) Apply risk-based technologies to high-risk significant components identified by the open-quotes expert panelclose quotes and outside of the IST program to determine whether additional testing requirements are appropriate

Risk-based inservice testing program modifications at Palo Verde nuclear generating station

Energy Technology Data Exchange (ETDEWEB)

Knauf, S.; Lindenlaub, B.; Linthicum, R.

1996-12-01

Arizona Public Service Company (APS) is investigating changes to the Palo Verde Inservice Testing (IST) Program that are intended to result in the reduction of the required test frequency for various valves in the American Society of Mechanical Engineers (ASME) Section XI IST program. The analytical techniques employed to select candidate valves and to demonstrate that these frequency reductions are acceptable are risk based. The results of the Palo Verde probabilistic risk assessment (PRA), updated in June 1994, and the risk significant determination performed as part of the implementation efforts for 10 CFR 50.65 (the maintenance rule) were used to select candidate valves for extended test intervals. Additional component level evaluations were conducted by an `expert panel.` The decision to pursue these changes was facilitated by the ASME Risk-Based Inservice Testing Research Task Force for which Palo Verde is participating as a pilot plant. The NRC`s increasing acceptance of cost beneficial licensing actions and risk-based submittals also provided incentive to seek these changes. Arizona Public Service is pursuing the risk-based IST program modification in order to reduce the unnecessary regulatory burden of the IST program through qualitative and quantitative analysis consistent with maintaining a high level of plant safety. The objectives of this project at Palo Verde are as follows: (1) Apply risk-based technologies to IST components to determine their risk significance (i.e., high or low). (2) Apply a combination of deterministic and risk-based methods to determine appropriate testing requirements for IST components including improvement of testing methods and frequency intervals for high-risk significant components. (3) Apply risk-based technologies to high-risk significant components identified by the {open_quotes}expert panel{close_quotes} and outside of the IST program to determine whether additional testing requirements are appropriate.

Developing the information management system for safeguards national inspection

International Nuclear Information System (INIS)

Park, S. J.; Jeon, I.; Park, W. S.; Min, K. S.

2003-01-01

The inspection information management system for safeguards national inspection is aimed to do the national safeguards inspection with efficiency, and to decrease the inspector's load to write inspection report by systematizing the inspection jobs and sharing the inspection data. National safeguards inspection is consisted two large jobs. The first is the national safeguards supporting job of managing to support the national inspection mission. The other is the writing a national inspection report after completing the national inspection. Before the developing of inspection information management system, the official tools(spread sheet, word processor) are usually used. But there is problem to share the data, to produce the statistics data. To solve the these problem, we developed the inspection information management system that process the job from initial to final inspection work, and opened user education. This paper explain the procedure of developing the inspection information management system for safeguards national inspection

Developing the information management system for safeguards national inspection

Energy Technology Data Exchange (ETDEWEB)

Park, S. J.; Jeon, I.; Park, W. S.; Min, K. S. [KAERI, Taejon (Korea, Republic of)

2003-10-01

The inspection information management system for safeguards national inspection is aimed to do the national safeguards inspection with efficiency, and to decrease the inspector's load to write inspection report by systematizing the inspection jobs and sharing the inspection data. National safeguards inspection is consisted two large jobs. The first is the national safeguards supporting job of managing to support the national inspection mission. The other is the writing a national inspection report after completing the national inspection. Before the developing of inspection information management system, the official tools(spread sheet, word processor) are usually used. But there is problem to share the data, to produce the statistics data. To solve the these problem, we developed the inspection information management system that process the job from initial to final inspection work, and opened user education. This paper explain the procedure of developing the inspection information management system for safeguards national inspection.

Status of ASME risk-based inspection guidelines development for nuclear power plants

International Nuclear Information System (INIS)

Gore, B.F.; Vo, T.V.; Balkey, K.R.

1991-01-01

The general methodology published by the task force is being applied to develop guidelines for the inspection of nuclear power plant components. The methodology has been expanded to include the use of information from probabilistic risk assessments to improve the quantification of risks associated with component ruptures. In addition, a procedure has been recommended for using these quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Finally, the method for determining the characteristics which an inspection strategy must possess in order to maintain these target rupture probabilities has been described. This paper presents the major features of this methodology, along with a discussion of the status of efforts to develop and to apply it further

A Bayesian analysis of component life expectancy and its implications on the inspection schedule

International Nuclear Information System (INIS)

Mason, Paolo

2017-01-01

A model of crack initiation and residual component life is fitted to the inspection history, inclusive of two in-service failures, of a set of gas circulator impellers at two UK power stations. The model is then used to estimate the probability of future in-service failure of each item in scenarios in which the next opportunity for inspection (i.e. detection of a developing crack) is exploited or forgone. The study takes into account in exact manner both variability and uncertainty. A novel Approximate Bayesian Computation (ABC) methodology is introduced for the quantification of uncertainty, i.e. the sampling of the posterior distribution of the parameters of the model. At the price of the discretization of this distribution, the methodology represents a quickly implemented option in problems where standard ABC rejection sampling is unacceptably inefficient. - Highlights: • A model of crack initiation and residual component life is introduced. • The model is fitted to the inspection log of a set of gas circulator impellers. • The odds of future failure are computed under different inspection schedules. • This prediction informs the choice to perform or forgo future inspection. • A novel Approximate Bayesian Computation algorithm is introduced and validated.

Study of a risk-based piping inspection guideline system.

Science.gov (United States)

Tien, Shiaw-Wen; Hwang, Wen-Tsung; Tsai, Chih-Hung

2007-02-01

A risk-based inspection system and a piping inspection guideline model were developed in this study. The research procedure consists of two parts--the building of a risk-based inspection model for piping and the construction of a risk-based piping inspection guideline model. Field visits at the plant were conducted to develop the risk-based inspection and strategic analysis system. A knowledge-based model had been built in accordance with international standards and local government regulations, and the rational unified process was applied for reducing the discrepancy in the development of the models. The models had been designed to analyze damage factors, damage models, and potential damage positions of piping in the petrochemical plants. The purpose of this study was to provide inspection-related personnel with the optimal planning tools for piping inspections, hence, to enable effective predictions of potential piping risks and to enhance the better degree of safety in plant operations that the petrochemical industries can be expected to achieve. A risk analysis was conducted on the piping system of a petrochemical plant. The outcome indicated that most of the risks resulted from a small number of pipelines.

A study on the optimization of test interval for check valves of Ulchin Unit 3 using the risk-informed in-service testing approach

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.; Yang, Z. A.; Ha, J. J.

2002-01-01

We optimized the test interval for check valves of Ulchin Unit 3 using the risk-informed in-service testing (IST) approach. First, we categorized the IST check valves for Ulchin Unit 3 according to their contributions to the safety of Ulchin Unit 3. Next, we performed the risk analysis on the relaxation of test interval for check valves identified as relatively low important to the safety of Ulchin Unit 3 to identify the maximum increasable test interval of them. Finally, we estimated the number of tests of IST check valves to be performed due to the changes of test interval. These study results are as follows: The categorization of IST check valve importance; the number of the HSSCs is 24(11.48%), the ISSCs is 40 (19.14%), and the LSSCs is 462(69.38%). The maximum increasable test interval; 6 times of current test interval of ISSCs2 and 40 times of that of LSSCs. The number of tests of IST check valves to be performed during 6 refueling time can be reduced from 7692 to 1333 ( 82.7%)

Boiler inspection manipulator for Torness Power Station

International Nuclear Information System (INIS)

Carrey, R.T.A.; Yule, I.Y.; Sibson, S.; Playle, M.J.

1996-01-01

The Advanced Gas-cooled Reactors at Torness and Heysham 2 are provided with dedicated access for remote inspection equipment. These in-service inspection (ISI) accesses comprise 12 penetrations above the core for inspection of the above core area and boilers, 12 below core penetrations for inspection of the lower boiler area and access through any of the 8 gas circulator penetrations for inspection of the sub-diagrid area. This paper describes a manipulator which will access the reactor from above the core via any of the 8 peripheral penetrations. (UK)

Use of expert judgment in the development and evaluation of risk-based inservice testing strategies for pumps and valves

International Nuclear Information System (INIS)

McAllister, W.J.; Perdue, R.K.; Balkey, K.R.; Closky, N.B.

1996-01-01

This paper describes a rigorous approach for quantitatively evaluating inservice testing effectiveness that evolved from two pilot plant studies. These studies prototyped methodologies for designing and selecting inservice testing (IST) strategies in a manner structured to insure that the targeted components will perform their required safety functions while minimizing life cycle inservice testing costs. The paper concentrates on the use of expert judgment in developing test effectiveness measures that move risk-based methods beyond ranking to optimization of plant IST programs. Selected results for check valves and pumps are shown to illustrate the practical significance of the approach

Use of expert judgment in the development and evaluation of risk-based inservice testing strategies for pumps and valves

Energy Technology Data Exchange (ETDEWEB)

McAllister, W.J.; Perdue, R.K.; Balkey, K.R.; Closky, N.B. [and others

1996-12-01

This paper describes a rigorous approach for quantitatively evaluating inservice testing effectiveness that evolved from two pilot plant studies. These studies prototyped methodologies for designing and selecting inservice testing (IST) strategies in a manner structured to insure that the targeted components will perform their required safety functions while minimizing life cycle inservice testing costs. The paper concentrates on the use of expert judgment in developing test effectiveness measures that move risk-based methods beyond ranking to optimization of plant IST programs. Selected results for check valves and pumps are shown to illustrate the practical significance of the approach.

Assessment of Current Inservice Inspection and Leak Monitoring Practices for Detecting Materials Degradation in Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Anderson, Michael T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Simonen, Fredric A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Muscara, Joseph [US Nuclear Regulatory Commission (NRC), Rockville, MD (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kupperman, David S. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-01

An assessment was performed to determine the effectiveness of existing inservice inspection (ISI) and leak monitoring techniques, and recommend improvements, as necessary, to the programs as currently performed for light water reactor (LWR) components. Information from nuclear power plant (NPP) aging studies and from the U. S. Nuclear Regulatory Commission’s Generic Aging Lessons Learned (GALL) report (NUREG-1801) was used to identify components that have already experienced, or are expected to experience, degradation. This report provides a discussion of the key aspects and parameters that constitute an effective ISI program and a discussion of the basis and background against which the effectiveness of the ISI and leak monitoring programs for timely detection of degradation was evaluated. Tables based on the GALL components were used to systematically guide the process, and table columns were included that contained the ISI requirements and effectiveness assessment. The information in the tables was analyzed using histograms to reduce the data and help identify any trends. The analysis shows that the overall effectiveness of the ISI programs is very similar for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). The evaluations conducted as part of this research showed that many ISI programs are not effective at detecting degradation before its extent reached 75% of the component wall thickness. This work should be considered as an assessment of NDE practices at this time; however, industry and regulatory activities are currently underway that will impact future effectiveness assessments. A number of actions have been identified to improve the current ISI programs so that degradation can be more reliably detected.

Auxiliary feedwater system risk-based inspection guide for the Palo Verde Nuclear Power Plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Sloan, J.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Palo Verde was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Palo Verde plants

Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1994-05-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant

Auxiliary feedwater system risk-based inspection guide for the Maine Yankee Nuclear Power Plant

International Nuclear Information System (INIS)

Gore, B.F.; Vo, T.V.; Moffitt, N.E.; Bumgardner, J.D.

1992-10-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. The information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Maine Yankee was selected as one of a series of plants for study. ne product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Maine Yankee plant

Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

International Nuclear Information System (INIS)

Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Vehec, T.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Point Beach was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRS. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Point Beach plant

Risk evaluation system for operational events and inspection findings

International Nuclear Information System (INIS)

Lopez G, A.; Godinez S, V.; Lopez M, R.

2010-10-01

The Mexican Nuclear Regulatory Commission has developed an adaptation of the US NRC Significance Determination Process (SDP) to evaluate the risk significance of operational events and inspection findings in Laguna Verde nuclear power plant. The Mexican Nuclear Regulatory Commission developed a plant specific flow chart for preliminary screening instead of the open questionnaire used by the US NRC-SDP, with the aim to improve the accuracy of the screening process. Also, the work sheets and support information tables required by the SDP were built up in an Excel application which allows to perform the risk evaluation in an automatic way, focusing the regulator staff efforts in the risk significance analysis instead of the risk calculation tasks. In order to construct this tool a simplified PRA model was developed and validated with the individual plant examination model. This paper shows the Mexican Nuclear Regulatory Commission process and some risk events evaluations performed using the Risk Evaluation System for Operational Events and Inspection Findings (SERHE, by its acronyms in Spanish). (Author)

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)

In-service quality assurance at the Fessenheim nuclear power plant

International Nuclear Information System (INIS)

Waller, A.

1980-01-01

Quality assurance is a combination of systematic and planned actions applied so as to give the adequate degree of confidence in the effective obtention of quality. Quality assurance is a daily requestioning resulting from systematic inspections and checks. The different principles permitting the in-service quality assurance system to be applied is presented [fr

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

Inspection of Nuclear Power Plant Containment Structures

Energy Technology Data Exchange (ETDEWEB)

Graves, H.L.; Naus, D.J.; Norris, W.E.

1998-12-01

Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1993-12-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant

Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

International Nuclear Information System (INIS)

Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1991-07-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab

Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

International Nuclear Information System (INIS)

Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V.; Garner, L.W.

1993-08-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant

On-line monitoring and inservice inspection in codes; Betriebsueberwachung und wiederkehrende Pruefungen in den Regelwerken

Energy Technology Data Exchange (ETDEWEB)

Bartonicek, J.; Zaiss, W. [Gemeinschaftskernkraftwerk Neckar GmbH, Neckarwestheim (Germany); Bath, H.R. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany). Geschaeftsstelle des Kerntechnischen Ausschusses (KTA)

1999-08-01

The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [Deutsch] Fuer die Komponentenintegritaet sind die Schaedigungsmechanismen mit dem nach den Regelwerken einzuhaltenden Abstand abzusichern. Dabei ist die jeweils vorhandene (Ist-) Qualitaet als Ausgangspunkt entscheidend. Die Absicherung der vorhandenen Qualitaet im weiteren Betrieb erfolgt durch geeignete Betriebsueberwachung und wiederkehrende Pruefungen. Die Anforderungen der Regelwerke sind vergleichbar, wobei die Bestimmung der vorhandenen Qualitaet nach einer bestimmten Betriebszeit sowie deren Absicherung im weiteren Betrieb am vollstaendigsten auf Basis des KTA-Regelwerkes moeglich ist. Die Absicherung der Komponentenintegritaet im Betrieb beruht in deutschen konventionellen Regelwerken nur auf den wiederkehrenden Pruefungen (hauptsaechlich Druckpruefungen und Sichtpruefungen). Das KTA-Regelwerk forderte hier schon immer qualifizierte

Qualification of NDT techniques for in-service inspections in nuclear power plants in accordance with ENIQ - examples and lessons - learned

International Nuclear Information System (INIS)

Just, T.; Csapo, G.

2006-01-01

ENIQ (European Network for Inspection and Qualification) has developed regulations on how to qualify non-destructive testing (NDT) methods and techniques in a standardized and structured manner. Two major innovative qualifications were carried out and reviewed with regard to implementation, according to the recommended German practice of ENIQ. The conclusions were drawn after performing the ENIQ qualification procedure for in-service inspections (ISI) of real components in nuclear power plants (NPP). The first example covers the qualification of NDT methods for the detection and characterization of surface, subsurface and underclad cracks in the area of the austenitic cladded RPV surface. Open and blind tests were conducted applying UT and ET (from the ID) and UT (from the OD) on realistic flaws (artificially induced IGSCC, hot cracks and fatigue cracks) in the cladding of a full scale RPV mock-up from MPA Stuttgart. The second example covers the qualification of mechanised RT in combination with tomography (developed by the BAM) for the sizing of cracks in pipe welds. For both qualification procedures TUEV NORD SysTec experts were part of the qualification body. The proposed NDT procedures have been qualified within defined limits of application. Recommendations were made to optimise the procedures and the techniques itself further. (orig.)

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

Evaluation of techniques for inspection and diagnostics of HWR pressure tubes

International Nuclear Information System (INIS)

Choi, Jong-Ho

2008-01-01

Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes for Heavy Water Reactors (HWRs), are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. The objective of the CRP was to inter-compare inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of the CRP, participants investigated the capability of different techniques to detect and characterize flaws. During the second phase, participants collaborated to determine the hydrogen concentration and to detect and characterize hydride blisters in zirconium alloy pressure tubes. Eight organizations from six countries, which operate HWRs, have participated in this CRP, Most of the techniques examined are well established and many of them are regularly used during in-service inspection of pressure tubes. The inter-comparison of these techniques provides a platform for identifying a particular technique (or a set of techniques), which is more accurate and reliable as compared to others for a specified task. The CRP also witnessed some new methodologies, which can be implemented on in-service inspection tools. These new techniques could complement the existing ones to overcome their limitations, thereby improving the reliability and accuracy of in-service inspection. This CRP also identified future areas of research and development. (author)

Discussion of in-service inspection planning for nuclear power plants

International Nuclear Information System (INIS)

Aoki, Takayuki; Takagi, Toshiyuki

2012-01-01

For creating an inspection plan for industrial plants like nuclear power plants, it is necessary to have a well-grounded basis. Therefore, this paper proposes a method for determining the following three elements of such a plan: the equipment to be inspected, the inspection method to be adopted, and the timing of its implementation using a scientific approach. The key idea is to analyze the components of the plan by employing the understanding of cancer detection and diagnosis in medical science and also to take an approach of creating a plan based on the characteristic features of the component and its aging degradation mode, the performance of the inspection method, and the relationship between the two. Taking all the above into account, the factors to be considered and the best way for inspecting the plants are presented in this paper. (author)

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

Auxiliary feedwater system risk-based inspection guide for the J.M. Farley Nuclear Power Plant

International Nuclear Information System (INIS)

Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G.

1990-10-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment(PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab

Investigating the Information Technology Courses for Pre-service and In-service English Teachers in Hong Kong

Directory of Open Access Journals (Sweden)

Chi Cheung Ruby Yang

2011-07-01

Full Text Available This paper focuses on the views of twenty-six pre-service and eight in-service English teachers in Hong Kong concerning the information technology courses in their teacher education programmes. The findings of the study show that while both software applications and technological integration in teaching are highlighted in the course outline, the instructor focused more on the application aspect, with the technology integration being less emphasised. Another important finding is that PowerPoint was still the most commonly used application in teaching. Likewise, usually the involved pre-service and in-service teachers just simply used different computer applications as alternatives to traditional teaching resources. This may reflect the fact that the information technology courses are not adequate to prepare teachers to teach with it.

Risk Management of Large RC Structures within Spatial Information System

DEFF Research Database (Denmark)

Qin, Jianjun; Faber, Michael Havbro

2012-01-01

Abstract: The present article addresses the development of a spatial information system (SIS), which aims to facilitate risk management of large‐scale concrete structures. The formulation of the SIS is based on ideas developed in the context of indicator‐based risk modeling for concrete structures...... subject to corrosion and geographical information system based risk modeling concerning large‐scale risk management. The term “risk management” here refers in particular to the process of condition assessment and optimization of the inspection and repair activities. The SIS facilitates the storage...... and handling of all relevant information to the risk management. The probabilistic modeling utilized in the condition assessment takes basis in a Bayesian hierarchical modeling philosophy. It facilitates the updating of risks as well as optimizing inspection plans whenever new information about the condition...

A practical approach for implementing risk-based inservice testing of pumps at nuclear power plants

International Nuclear Information System (INIS)

Hartley, R.S.; Maret, D.; Seniuk, P.; Smith, L.

1996-01-01

The American Society of Mechanical Engineers (ASME) Center for Research and Technology Development's (CRTD) Research Task Force on Risk-Based Inservice Testing has developed guidelines for risk-based inservice testing (IST) of pumps and valves. These guidelines are intended to help the ASME Operation and Maintenance (OM) Committee to enhance plant safety while focussing appropriate testing resources on critical components. This paper describes a practical approach for implementing those guidelines for pumps at nuclear power plants. The approach, as described in this paper, relies on input, direction, and assistance from several entities such as the ASME Code Committees, United States Nuclear Regulatory Commission (NRC), and the National Laboratories, as well as industry groups and personnel with applicable expertise. Key parts of the risk-based IST process that are addressed here include: identification of important failure modes, identification of significant failure causes, assessing the effectiveness of testing and maintenance activities, development of alternative testing and maintenance strategies, and assessing the effectiveness of alternative testing strategies with present ASME Code requirements. Finally, the paper suggests a method of implementing this process into the ASME OM Code for pump testing

A practical approach for implementing risk-based inservice testing of pumps at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Hartley, R.S. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Maret, D.; Seniuk, P.; Smith, L.

1996-12-01

The American Society of Mechanical Engineers (ASME) Center for Research and Technology Development`s (CRTD) Research Task Force on Risk-Based Inservice Testing has developed guidelines for risk-based inservice testing (IST) of pumps and valves. These guidelines are intended to help the ASME Operation and Maintenance (OM) Committee to enhance plant safety while focussing appropriate testing resources on critical components. This paper describes a practical approach for implementing those guidelines for pumps at nuclear power plants. The approach, as described in this paper, relies on input, direction, and assistance from several entities such as the ASME Code Committees, United States Nuclear Regulatory Commission (NRC), and the National Laboratories, as well as industry groups and personnel with applicable expertise. Key parts of the risk-based IST process that are addressed here include: identification of important failure modes, identification of significant failure causes, assessing the effectiveness of testing and maintenance activities, development of alternative testing and maintenance strategies, and assessing the effectiveness of alternative testing strategies with present ASME Code requirements. Finally, the paper suggests a method of implementing this process into the ASME OM Code for pump testing.

Framework for applying RI-ISI methodology for Indian PHWRs

International Nuclear Information System (INIS)

Vinod, Gopika; Saraf, R.K.; Ghosh, A.K.; Kushwaha, H.S.

2006-01-01

Risk Informed In-Service Inspection (RI-ISI) aims at categorizing the components for In-Service inspection based on their contribution to Risk. For defining the contribution of risk from components, their failure probabilities and its subsequent effect on Core Damage Frequency (CDF) needs to be evaluated using Probabilistic Safety Assessment methodology. During the last several years, both the U.S. Nuclear Regulatory Commission (NRC) and the nuclear industry have recognized that Probabilistic Safety Assessment (PSA) has evolved to be more useful in supplementing traditional engineering approaches in reactor regulation. The paper highlights the various stages involved in applying RI-ISI and then compares the findings with existing ISI practices. (author)

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

The application of the PSA important measures in risk-informed administrations

International Nuclear Information System (INIS)

Chen Yan; Fu Zhiwei; Jing Jianping; Zhang Chunming; Liu Hongquan

2012-01-01

The importance measures analyses of PSA are main approaches during the risk-informed administrations. This paper reviews kinds of importance measures, mainly researches the meaning of the FV and RAW importance measures, and introduces the applications of importance measures in the in-service testing and categorization of SSCs, finally, discusses the limitations of the importance measures analyses. (authors)

The role of NDE in maintaining and extending the life cycle of engineering components

International Nuclear Information System (INIS)

Doctor, S.R.

1997-01-01

A successful life cycle of engineering components begins during the design to select the best materials for a given application, designing for economical maintenance, and accommodating NDE inspections. NDE in the nuclear industry is part of the defense in depth to insure structural integrity of components. High reliability NDE is needed to reliably detect and accurately characterize the failure modes that have occurred in nuclear components. The move toward risk-informed inservice inspection programs focusses the NDE on the most safety significant components. The use of performance demonstration testing is leading to improved inservice inspections through screening out ineffective procedure/equipment/personnel. (orig.)

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.)

Risk assessment application to NRC inspection

International Nuclear Information System (INIS)

Campbell, D.J.; Guthrie, V.H.; Flanagan, G.F.

1987-01-01

Inspectors must make many decisions on the allocation of their efforts. To date, these decisions have been made based upon their own judgment and guidance from inspection procedures. The program described in this paper provides PRA information as an additional aid to inspectors. A structured approach for relating PRA information to specific inspection decisions has been developed. The use of PRA information as an aid in optimal decision making (1) in response to the current plant status and (2) in the scheduling of effort over an extended period of time is considered. (orig.)

Understanding household preferences for hurricane risk mitigation information: evidence from survey responses.

Science.gov (United States)

Chatterjee, Chiradip; Mozumder, Pallab

2014-06-01

Risk information is critical to adopting mitigation measures, and seeking risk information is influenced by a variety of factors. An essential component of the recently adopted My Safe Florida Home (MSFH) program by the State of Florida is to provide homeowners with pertinent risk information to facilitate hurricane risk mitigation activities. We develop an analytical framework to understand household preferences for hurricane risk mitigation information through allowing an intensive home inspection. An empirical analysis is used to identify major drivers of household preferences to receive personalized information regarding recommended hurricane risk mitigation measures. A variety of empirical specifications show that households with home insurance, prior experience with damages, and with a higher sense of vulnerability to be affected by hurricanes are more likely to allow inspection to seek information. However, households with more members living in the home and households who live in manufactured/mobile homes are less likely to allow inspection. While findings imply MSFH program's ability to link incentives offered by private and public agencies in promoting mitigation, households that face a disproportionately higher level of risk can get priority to make the program more effective. © 2014 Society for Risk Analysis.

Piping inspection round robin

International Nuclear Information System (INIS)

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

1996-04-01

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

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

49 CFR 556.9 - Public inspection of relevant information.

Science.gov (United States)

2010-10-01

... 49 Transportation 6 2010-10-01 2010-10-01 false Public inspection of relevant information. 556.9... NONCOMPLIANCE § 556.9 Public inspection of relevant information. Information relevant to a petition under this... Administration, 400 Seventh Street, SW., Washington, DC 20590. Copies of available information may be obtained in...

PWR reactor vessel in-service-inspection according to RSEM

International Nuclear Information System (INIS)

Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

2006-01-01

Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

Feasibility of risk-informed technology for japanese nuclear power plants

International Nuclear Information System (INIS)

Yoshida, Tomoo; Fujioka, Terutaka; Kirimoto, Yorihiro; Ueda, Nobuyuki; Kinoshita, Izumi; Kashima, Koichi

2000-01-01

Risk-informed technology utilizes Probabilistic Safety Assessment for streamlining the maintenance of nuclear power plants. With this technology, plant components are categorized as either high or low-safety-significant components. The Maintenance requirements focuses on high safety-significant components and are relieved for low safety significant ones. This is expected to reduce plant cost while maintaining safety. We investigated especially risk-informed inservice inspection of piping in U.S. nuclear power plants in the interest of determining its feasibility for Japanese plants. Quantitative and qualitative RI-ISI methods were developed by the ASME/Westinghouse Owners Group and EPRI, respectively. These methods have been incorporated in the ASME Section 11 Code Cases and endorsed by the U.S. Nuclear Regulatory Commission. The quantitative method evaluates component segment risks in terms of pipe failure probability calculated with a probabilistic fracture mechanics(PFM) model and pipe failure impact categorization on core damage frequency(CDF) calculated with PSA. The qualitative method uses pipe failure potential categorization derived from the plant service experiences and pipe failure impact on CDF derived from the PSA insight. The PFM model is applicable only to failures from initial welding defects and stress corrosion cracking, therefore it does not cover such significant failure mechanisms found in nuclear power plants as corrosion or high-cycle fatigue, etc.. Thus, a qualitative failure potential categorization method was developed on the basis of the service experiences of the U.S. nuclear power plants, so that appropriate categorization rules must be developed on the service experiences in Japanese plants. Accordingly, we have devised a software framework with a computer-aided system for the selection of risk significant elements. This system consists of a piping failure database module, a piping failure analysis module, and a piping failure potential

Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, Units 1 and 2 nuclear power plants

International Nuclear Information System (INIS)

Lloyd, R.C.; Vehec, T.A.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Rossbach, L.W.; Sena, P.P. III

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Beaver Valley Units 1 and 2 were selected as two of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at Beaver Valley Units 1 and 2

Manipulator arm for a nuclear reactor vessel inspection device

International Nuclear Information System (INIS)

1980-01-01

A manipulator arm for a reactor vessel in-service inspection apparatus is adapted to transport a transducer array for ultrasonic examination of welds at any point in the vessel. The removal of the inspection device from the reactor vessel in an emergency presents a problem where a relatively long manipulator arm is used. This invention provides an improved arm with means for changing the normal orientation of the arm to a shorter one to permit safe removal of the inspection device from the reactor vessel. (author)

A web environment for inspection service; Un entorno WEB para la inspeccion en servicio

Energy Technology Data Exchange (ETDEWEB)

Robles Guemes, N.; Garcia Heras, M.; Cueto-Felgueroso Garcia, C.; Pelaez Gutierrez, J. A.

2013-07-01

The in-service inspection fundamental objective is to maintain the structural integrity of structures, systems and components during the life in the installation service. Nuclear power plants are required to implement a Service inspection program in accordance with the established regulation by the Regulatory body.

Guided wave technology for in-service inspection and online monitoring for long term operation of nuclear power plants

International Nuclear Information System (INIS)

Bertoncini, Francesco; Raugi, Marco; Cappelli, Mauro; Cordella, Francesco; Mazzini, Davide

2015-01-01

In-Service Inspection (ISI) and monitoring of all equipment (Systems, Structures and Components, SSCs) of a Nuclear Power Plant (NPP), are actions aimed at preventing failures both for economical and safety purposes. SSCs ageing due to stresses such as corrosion, load variations, flow conditions, temperature and neutron irradiation can be a potential limit for NPP life extension or operation beyond their license term (Long Term Operation. LTO). ISI has a main role on the actual possibility of LTO assuring the required safety. Guided Waves are structure-borne ultrasonic waves that propagate along the structure confined and guided by its geometric boundaries. Guided Wave Testing can find defect locations through long-range screening using low-frequency waves (from 5 to 250 kHz). The technology is regularly used for pipe testing in the oil and gas industry. In the nuclear industry, regulators are working to standardize monitoring and inspection procedures. To use the technology inside an active plant, operators must solve issues like high temperatures (up to more than 300degC inside a light-water reactor's primary piping), high wall thickness of components in the primary circuit and characteristic defect typologies. Magnetostrictive sensors are expected to overcome such issues due to their physical properties, namely robust constitution and simplicity. Recent experimental results have demonstrated magnetostrictive transducers can withstand temperatures close to 300degC. In this paper, new experimental tests conducted using such a methodology will be described and open issues related to high temperature guided wave applications (e.g. wave velocity or amplitude fluctuations during propagation in variable temperature components) will be discussed. (author)

Development of the Risk-Based Inspection Techniques and Pilot Plant Activities

International Nuclear Information System (INIS)

Phillips, J.H.

1997-01-01

Risk-based techniques have been developed for commercial nuclear power plants. System boundaries and success criteria is defined using the probabilistic risk analysis or probabilistic safety analysis developed to meet the individual plant evaluation. Final ranking of components is by a plant expert panel similar to the one developed for maintenance rule. Components are identified as being high risk-significant or low-risk significant. Maintenance and resources are focused on those components that have the highest risk-significance. The techniques have been developed and applied at a number of pilot plants. Results from the first risk-based inspection pilot plant indicates that safety due to pipe failure can be doubled while the inspection reduced to about 80% when compared with current inspection programs. The reduction in inspection reduces the person-rem exposure resulting in further increases in safety. These techniques have been documented in publication by the ASME CRTD

Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

Energy Technology Data Exchange (ETDEWEB)

Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P. [and others

1996-12-01

This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising.

Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

International Nuclear Information System (INIS)

Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P.

1996-01-01

This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising

Boiler inspection manipulator for Torness Power Station

International Nuclear Information System (INIS)

Carrey, R.T.A.; Playle, M.J.

1996-01-01

In-service inspection at Torness Advanced Gas-cooled Reactor is achieved, via penetrations above and below the core area and boilers and through eight circulator penetrations for inspection of the sub-diagrid area. A manipulator is described which can access the area above the reactor core via any of these eight peripheral penetrations. The design and fabrication of the manipulator has led to innovation with a number of possible solutions being tendered. In reactor deployment of the successful design is expected in early 1997. (UK)

Ultrasonic inspection of inpile tubes

International Nuclear Information System (INIS)

Boyd, D.M.; Bossi, H.

1985-01-01

The in-service inspection (ISI) of inpile tubes can be performed accurately and safely with a semiautomatic ultrasonic inspection system. The ultrasonic technique uses a set of multiple transducers to detect and size cracks, voids, and laminations radially and circumferentially. Welds are also inspected for defects. The system is designed to inspect stainless steel and Inconel tubes ranging from 53.8 mm (2.12 in.) to 101.6 mm (4 in.) inner diameter with wall thickness on the order of 5 mm. The inspection head contains seven transducers mounted in a surface-following device. Six angle-beam transducers generate shear waves in the tubes. Two of the six are oriented to detect circumferential cracks, and two detect axial cracks. Although each of these four transducers is used in the pulse-echo mode, they are oriented in aligned sets so pitch-catch operation is possible if desired. The remaining angle-beam transducers are angulated to detect flaws that are off axial or circumferential orientation. The seventh transducer is used for longitudinal inspection and detects and sizes laminar-type defects

Risk based inspection for atmospheric storage tank

Science.gov (United States)

Nugroho, Agus; Haryadi, Gunawan Dwi; Ismail, Rifky; Kim, Seon Jin

2016-04-01

Corrosion is an attack that occurs on a metallic material as a result of environment's reaction.Thus, it causes atmospheric storage tank's leakage, material loss, environmental pollution, equipment failure and affects the age of process equipment then finally financial damage. Corrosion risk measurement becomesa vital part of Asset Management at the plant for operating any aging asset.This paper provides six case studies dealing with high speed diesel atmospheric storage tank parts at a power plant. A summary of the basic principles and procedures of corrosion risk analysis and RBI applicable to the Process Industries were discussed prior to the study. Semi quantitative method based onAPI 58I Base-Resource Document was employed. The risk associated with corrosion on the equipment in terms of its likelihood and its consequences were discussed. The corrosion risk analysis outcome used to formulate Risk Based Inspection (RBI) method that should be a part of the atmospheric storage tank operation at the plant. RBI gives more concern to inspection resources which are mostly on `High Risk' and `Medium Risk' criteria and less on `Low Risk' shell. Risk categories of the evaluated equipment were illustrated through case study analysis outcome.

Mechanized inspection of steam generator components during manufacture

International Nuclear Information System (INIS)

Otte, H.-J.; Leupoldt, K.; Meister, W.

2009-01-01

Steam Generator (SG) parts are intensively inspected by UT in the course of the manufacturing process. These inspections - mostly performed manually using different codes - are time consuming and call for a sophisticated documentation, figuring part of the life time documentation package. In order to reduce time and costs mechanized inspection equipment is introduced, combining short inspection times, avoiding influence of the human factor and providing proper electronic storage of all inspection results prepared for comparison with data generated during in-service inspection. Since 2001 Cegelec delivered various UT systems for gas turbine disks and rotor ends called SIRO-MAN. Within only a few years the majority of important providers of such components successfully switched from manual inspection to mechanized inspection following the requirements of manufacturers like ALSTOM, GE and Siemens. The SIRO-MAN is now adapted to the needs of mechanized inspection of SG components. The inspection is performed on the products during rotation around the vertical axis. The multi - probe assemblies are manoeuvred on the products by a manipulator system backed by a NC control unit. Acoustic coupling of UT probes to the product surface is performed with oil or water in a closed circuit. UT and - if requested ET - data along with position information of the probe assembly provided by the control unit are acquired, processed and evaluated by an UT / ET electronic system delivered by either Olympus or ZETEC. As performed already on rotor ends a sequence of inspections using different parameter settings can be programmed with simple means (Teach In) so that such inspection sequence can be executed without operating personnel. Probe assemblies allow for individual operation of probes out of the probe assembly according to the individual needs. Conventional UT and phased array applications or combination of both techniques can be provided. The UT / ET electronic equipment offers

In-service inspection and maintenance schedule for a typical TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Boeck, H.

1996-05-01

This report lists all the systems and components of the TRIGA reactor Vienna which are inspected and maintained in regular intervals. These intervals are categorized in monthly, quarterly, semi-annual and annual inspections. Further the type of inspection and the responsibility for the inspection is shown. For each component specific inspection sheets have been developed, some examples are given in the annex. (author)

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

Ultrasonic inspection development at HEDL

International Nuclear Information System (INIS)

Day, C.K.; Mech, S.J.; Michaels, T.E.; Dixon, N.E.

1978-01-01

Ultrasonic testing methods and equipment are being developed to support preservice and in-service inspection of selected FFTF welds. A digital computer system is employed in the analysis of both simulated FFTF pipe sections and plate specimens containing fatigue cracks. It is anticipated that test evaluation standards containing fatigue cracks will partially eliminate questions formerly associated with weld test calibration producers by providing natural cracks which follow grain boundaries and stress patterns resembling piping situ conditions. Studies have revealed that commercial transducers may satisfy LMFBR ultrasonic pipe inspection applications: The test system evaluation included transducers and wedge coupling and fluid coupling materials which exhibited acceptable performance at temperatures to 2300C. Results are presented that demonstrate the feasibility of ultrasonic inspection of components immersed in sodium at temperatures to 2600C. (UK)

Risk-Based School Inspections: Impact of Targeted Inspection Approaches on Dutch Secondary Schools

Science.gov (United States)

Ehren, Melanie C.; Shackleton, Nichola

2016-01-01

In most countries, publicly funded schools are held accountable to one inspectorate and are judged against agreed national standards. Many inspectorates of education have recently moved towards more proportional risk-based inspection models, targeting high-risk schools for visits, while schools with satisfactory student attainment levels are…

Code cases for implementing risk-based inservice testing in the ASME OM code

International Nuclear Information System (INIS)

Rowley, C.W.

1996-01-01

Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices

Code cases for implementing risk-based inservice testing in the ASME OM code

Energy Technology Data Exchange (ETDEWEB)

Rowley, C.W.

1996-12-01

Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices.

Management of Microbiologically Influenced Corrosion in Risk-Based Inspection Analysis

DEFF Research Database (Denmark)

Skovhus, Torben Lund; Andersen, Erlend S.; Hillier, Elisabeth

2018-01-01

Operating offshore oil and gas production facilities is often associated with high risk. To manage the risk, operators commonly use aids to support decision making in the establishment of a maintenance and inspection strategy. Risk-based inspection (RBI) analysis is widely used in the offshore...... industry as a means to justify the inspection strategy adopted. The RBI analysis is a decision-making technique that enables asset managers to identify the risk related to failure of their most critical systems and components, with an effect on safety, environmental, and business-related issues. Risk...... influenced corrosion (MIC) is a degradation mechanism that has received increased attention from corrosion engineers and asset operators in the past decades. In this paper, the most recent models that have been developed to assess the impact of MIC on asset integrity will be presented and discussed. From...

Inspection qualification as a tool to risk based ET ISI of VVER type SG tubes

International Nuclear Information System (INIS)

Horacek, L.

2002-01-01

A Pilot study on Eddy current inspection qualification of VVER 440 steam generator tubes, discussed in this paper, followed the ENIQ methodology principles and covered briefly the assumed scope of ET qualification, relevant elaborated qualification documents, known ISI limitations and a review of input information on component and defects determined for Eddy current inspection qualification of VVER 440 steam generator tubes. The information includes the fabrication of the test blocks with SG tube segments provided by intended defect simulations of realistic SCC type and basic data on the realistic SCC type defects manufacturing technology. Lessons learned from the development of manufacturing technology of SSC type of defects, regional blind tests, elaboration of the preliminary technical justification for Eddy current automated inspections, potential optimisation of inspection procedures, laboratory and practical open trials are summarised in the paper. The results of the Pilot study also especially in relation to POD curve being determined seem to be useful for practical operational ISI programme and Risk informed ISI decisions and the establishment of plugging criteria of VVER 440 and VVER 1000 type steam generator tubes. (orig.)

Review of inservice inspections of greased tendons in prestressed-concrete containments

International Nuclear Information System (INIS)

Dougan, J.R.; Ashar, H.

1983-01-01

Prestressed-concrete containments in the United States using greased prestressing tendons are inspected periodically to ensure structural integrity and to identify and correct problem areas before they become critical. An analysis of the available utility inspection data and an evaluation of the current and proposed guidelines were conducted to provide a measure of the reliability of the inspection process. Comments from utility and industry personnel were factored into the analysis. The results indicated that the majority of the few incidences of problems or abnormalities which occurred were minor in nature and did not threaten the structural integrity of the containment

An experience with in-service fabrication and inspection of austenitic stainless steel piping in high temperature sodium system

Energy Technology Data Exchange (ETDEWEB)

Ravi, S., E-mail: sravi@igcar.gov.in; Laha, K.; Sakthy, S.; Mathew, M.D.; Bhaduri, A.K.

2015-04-01

Highlights: • Procedure for changing 304L SS pipe to 316L SS in sodium loop has been established. • Hot leg made of 304L SS was isolated from existing cold leg made of 316LN SS. • Innovative welding was used in joining the new 316L SS pipe with existing 316LN SS. • The old components of 304L SS piping have been integrated with the new piping. - Abstract: A creep testing facility along with dynamic sodium loop was installed at Indira Gandhi Centre for Atomic Research, Kalpakkam, India to assess the creep behavior of fast reactor structural materials in flowing sodium. Type 304L austenitic stainless steel was used in the low cross section piping of hot-leg whereas 316LN austenitic stainless steel in the high cross section cold-leg of the sodium loop. The intended service life of the sodium loop was 10 years. The loop has performed successfully in the stipulated time period. To enhance its life time, it has been decided to replace the 304L piping with 316L piping in the hot-leg. There were more than 300 welding joints involved in the integration of cold-leg with the new 316L hot-leg. Continuous argon gas flow was maintained in the loop during welding to avoid contamination of sodium residue with air. Several innovative welding procedures have been adopted for joining the new hot-leg with the existing cold-leg in the presence of sodium residue adopting TIG welding technique. The joints were inspected for 100% X-ray radiography and qualified by performing tensile tests. The components used in the discarded hot-leg were retrieved, cleaned and integrated in the renovated loop. A method of cleaning component of sodium residue has been established. This paper highlights the in-service fabrication and inspection of the renovation.

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.

Generic Reliability-Based Inspection Planning for Fatigue Sensitive Details

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Straub, Daniel; Faber, Michael Havbro

2005-01-01

of fatigue sensitive details in fixed offshore steel jacket platforms and FPSO ship structures. Inspection and maintenance activities are planned such that code based requirements to the safety of personnel and environment for the considered structure are fulfilled and at the same time such that the overall......The generic approach for planning of in-service NDT inspections is extended to cover the case where the fatigue load is modified during the design lifetime of the structure. Generic reliability-based inspection planning has been developed as a practical approach to perform inspection planning...... expected costs for design, inspections, repairs and failures are minimized. The method is based on the assumption of “no-finds” of cracks during inspections. Each fatigue sensitive detail is categorized according to their type of details (SN curves), FDF values, RSR values, inspection, repair and failure...

Opinion on the demonstration of the 900 MWe reactor vessels in-service behaviour after their third decennial inspection

International Nuclear Information System (INIS)

2010-01-01

In this report, an expert group comments and assesses how sufficient are the demonstration and the actions performed by EDF to justify the in-service behaviour of nuclear reactor vessels. More precisely, it comments and discusses the different steps of the EDF demonstration: follow-up of the fluence received by the vessels, identification of the most severe transients and thermodynamic calculations, behaviour of irradiated materials, mechanical analysis, in-service control and follow-up plan, ageing management. Recommendations are then formulated

Potential for ultrasonic inspection of heat exchanger tubes

International Nuclear Information System (INIS)

Ward, M.J.

1980-01-01

Preliminary results of a program to develop an ultrasonic inspection method for heat-exchanger tubes, to be used as alternative or complementary to eddy-curent testing were sufficiently promising to warrant further study. Problems were encountered in adapting a standard commercial transducer and a custom-made transducer to provide full 360 degree coverage of an area for in-service inspection, but it might be possible to overcome these problems. The results showed it might prove impossible to design a transducer to handle the tight U-bends in some heat exchangers. The most promising area of application for ultrasonic inspection was found to be around the tubesheet. (DN)

In-service examination of IHX tubing with eddy current NDT equipment

International Nuclear Information System (INIS)

Brown, R.L.

1972-01-01

Single and multiple frequency eddy current (ET) nondestructive testing (NDT) techniques and equipment were investigated for in-service inspection of sodium-contaminated intermediate heat exchanger (IHX) tubing. A four frequency technique, demonstrated in the laboratory, was relatively insensitive to signals caused by probe motion, tube support plates, and residual sodium on the outer surface of the tubes. No method was found to avoid the signals from residual sodium on the inside surfaces of the tube. (U.S.)

Risk-informed prediction of feeder end of life

International Nuclear Information System (INIS)

Jyrkama, M.; Pandey, M.

2011-01-01

The operating life of feeder piping is negatively impacted by flow accelerated corrosion (FAC). In this study, an assessment of a large set of inspection data reveals that FAC in feeders is a relatively stationary process, with variability only at the local scale. Given the added uncertainty from inspection coverage, a new method for estimating the thinning rate and feeder EOL is developed using a probabilistic approach. The results of the study illustrate the benefits of the methodology in supporting risk-informed decision making at the station by quantifying the present and incremental risk in the feeder system over time. (author)

Risk-informed prediction of feeder end of life

Energy Technology Data Exchange (ETDEWEB)

Jyrkama, M.; Pandey, M. [Univ. of Waterloo, Ontario (Canada)

2011-07-01

The operating life of feeder piping is negatively impacted by flow accelerated corrosion (FAC). In this study, an assessment of a large set of inspection data reveals that FAC in feeders is a relatively stationary process, with variability only at the local scale. Given the added uncertainty from inspection coverage, a new method for estimating the thinning rate and feeder EOL is developed using a probabilistic approach. The results of the study illustrate the benefits of the methodology in supporting risk-informed decision making at the station by quantifying the present and incremental risk in the feeder system over time. (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)

Pressure vessel integrity and weld inspection procedure

International Nuclear Information System (INIS)

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

1975-01-01

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

Benefits of Risk Based Inspection Planning for Offshore Structures

DEFF Research Database (Denmark)

Straub, D.M.; Goyet, J.; Sørensen, John Dalsgaard

2006-01-01

The economical benefits of applying risk-based inspection planning (RBI) for offshore structures subject to fatigue are evaluated based on experiences from past industrial projects. To this end, the factors influencing the cost of inspection, repair and failure of structures are discussed......, the financial benefit of RBI is assessed....

Alternate Welding Processes for In-Service Welding

Science.gov (United States)

2009-04-24

Conducting weld repairs and attaching hot tap tees onto pressurized pipes has the advantage of avoiding loss of service and revenue. However, the risks involved with in-service welding need to be managed by ensuring that welding is performed in a rep...

Information managing in 60Co container inspection system

International Nuclear Information System (INIS)

Wu Zhifang; Gu Bohua; Zhou Liye; An Jigang; Liu Yisi

1998-01-01

The design, maintenance and realization of information managing database in 60 Co container inspection system made by INET of Tsinghua University is introduced. The technique of Open Database Connectivity (ODBC) is adopted to develop a general format database including text and graphic information. The database application is developed with Visual C ++ 5.0 programming language to run in 32-bit Windows operation system. It conforms to Client/Server model and supports network communication. It works very well in the laboratory emulator of 60 Co container inspection system

Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

International Nuclear Information System (INIS)

Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

2006-02-01

The primary components such as a reactor core, heat exchangers, pumps and internal structures 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 and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A 0 plate wave was selected as the application mode of the sensor. The A 0 plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A 0 mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote inspection

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)

Ferromagnetic material inspection for feedwater heater and condenser tubes

International Nuclear Information System (INIS)

Anon.

1987-01-01

In recent years, special ferritic stainless steels, such as AL29-4C/sup TM/, Sea-Cure/sup TM/, E-Brite/sup TM/, 439, and similar alloys have been introduced as tube material in condensers, feedwater heaters, moisture separator/reheaters, and other heat exchangers. In addition, carbon steel tubes are widely used in feedwater heaters and heat exchangers in chemical plants. The main problem with the in-service inspection of these ferritic alloys and carbon steel tubes lies in their highly ferromagnetic properties. These properties severely limit the application of the standard eddy current techniques. The effort was undertaken under EPRI sponsorship to develop a reliable technique for in-service inspection of ferromagnetic tubes. The new method combines the measurement of magnetic flux leakage generated around the defects with measurement of total flux in the tube wall. The heart of the inspection system is a special ID probe that magnetizes the tube and generates signals for any tube defect. A permanent record of inspection is provided with a strip-chart or magnetic tape recorder. The laboratory and field evaluation of this new system demonstrated its very good sensitivity to small defects, its reliability, and its ruggedness. Defects as small as 10% external wall loss in heavy wall carbon steel tube were detected. Tubes in the power plant were inspected at a rate of 300-500 tubes per eight-hour shift. The other advantages of this newly developed technique are its simplicity, low cost of instrumentation, easy data interpretation, and full portability

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

Guidelines for inservice testing at nuclear power plants

International Nuclear Information System (INIS)

Campbell, P.

1995-04-01

The staff of the U.S. Nuclear Regulatory Commission (NRC) gives licensees guidelines and recommendations for developing and implementing programs for the inservice testing of pumps and valves at commercial nuclear power plants. The staff discusses the regulations; the components to be included in an inservice testing program; and the preparation and content of cold shutdown justifications, refueling outage justifications, and requests for relief from the American Society of Mechanical Engineers Code requirements. The staff also gives specific guidance on relief acceptable to the NRC and advises licensees in the use of this information at their facilities. The staff discusses the revised standard technical specifications for the inservice testing program requirements and gives guidance on the process a licensee may follow upon finding an instance of noncompliance with the Code

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.

Summary report for 1990 inservice inspection (ISI) of SRS 100-L reactor tank

International Nuclear Information System (INIS)

Morrison, J.M.; Loibl, M.W.

1991-01-01

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in about 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. The primary objective of this inspection was to determine if the accessible welds and selected portions of base metal in the L Reactor tank wall contain any indications of IGSCC. This inspection included areas in and beyond the weld HAZ, extending out as far as two to three inches from the centerline of the welds, plus selected areas of base metal at the intersection of the main tank vertical and mid-girth welds. No evidence of such degradation was found in any of the areas examined. Further, additional inspections were conducted of areas that had been damaged and repaired during original fabrication, and on a sample of areas containing linear indications observed during the 1986 visual inspection of the tank. No evidence of IGSCC or other service induced degradation was detected in these areas, either. The inspection was initially planned to cover a minimum of 60% of the accessible welds, plus repair areas and a sample of the indications from the 1986 visual inspection. Direction was received from DOE while the inspection was in progress to expand the scope to cover 100% of the accessible weld areas, and the plan was adjusted accordingly. Initial setup of the tank, which prior to inspection contained Mark 60B target assemblies and nearly a full charge of Mark 22 fuel assemblies, began on October 15, 1990. The inspection was completed on April 12, 1991

Risk informed life cycle plant design

International Nuclear Information System (INIS)

Hill, Ralph S. III; Nutt, Mark M.

2003-01-01

Many facility life cycle activities including design, construction, fabrication, inspection and maintenance are evolving from a deterministic to a risk-informed basis. The risk informed approach uses probabilistic methods to evaluate the contribution of individual system components to total system performance. Total system performance considers both safety and cost considerations including system failure, reliability, and availability. By necessity, a risk-informed approach considers both the component's life cycle and the life cycle of the system. In the nuclear industry, risk-informed approaches, namely probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA), have become a standard tool used to evaluate the safety of nuclear power plants. Recent studies pertaining to advanced reactor development have indicated that these new power plants must provide enhanced safety over existing nuclear facilities and be cost-competitive with other energy sources. Risk-informed approaches, beyond traditional PRA, offer the opportunity to optimize design while considering the total life cycle of the plant in order to realize these goals. The use of risk-informed design approaches in the nuclear industry is only beginning, with recent promulgation of risk-informed regulations and proposals for risk-informed codes. This paper briefly summarizes the current state of affairs regarding the use of risk-informed approaches in design. Key points to fully realize the benefit of applying a risk-informed approach to nuclear power plant design are then presented. These points are equally applicable to non-nuclear facilities where optimization for cost competitiveness and/or safety is desired. (author)

Unique systems for inspection and repair of VVER steam generators

International Nuclear Information System (INIS)

Kovalyk, Adrian; Pilat, Peter; Jablonicky, Pavol

2014-01-01

During over 30 years, VUJE Trnava has developed a series of remote control manipulators for in-service inspection and maintenance of steam generator collectors. The manipulators have been widely used on WWER type reactors in Slovakia and abroad. Some of the manipulators for non-destructive testing are shown.

Bridging of inspection with corrosion management

International Nuclear Information System (INIS)

Zamaluddin Ali; Mohd Hawari Hassan; Rohana Jaafar

2009-01-01

Formerly, Inspection and Corrosion Engineers have less interaction and sharing of information to each other even they are working in the same plant or organization. Inspection finding either from visual inspection or NDT techniques rarely shared with corrosion engineers. Similarly corrosion engineers rarely discussed their corrosion prediction and potential damage mechanism with inspection engineers. A demanding request of more holistic plant safety and asset integrity promoted the introduction and implementation of Risk Based Inspection (RBI). RBI analysis demands the input mainly from both disciplines i.e. Inspection and Corrosion Engineers. Most of RBI methodologies are once-off analysis approach which also promoted once-off interaction between Inspection and Corrosion Engineers. PETRONAS has developed a methodology with supporting software, integrating both Inspection and Corrosion disciplines. PETRONAS Risk Based Inspection (PRBI) is intended to promote continuous integration of Inspection and Corrosion management of the plant through out the whole life cycle starting from the design stage to fabrication, operation and decommissions stage. (author)

Experiences and future plan for risk-informed decision making in Korea

International Nuclear Information System (INIS)

Chung, D.W.; Ryu, Y.H.

2001-01-01

A program for establishing regulatory framework on the use of risk information has been underway since 1995, and several trial applications have been done to evaluate the applicability and usefulness of risk-informed approach to nuclear power plant regulation. The program consists of fifteen general and/or specific items of interest and pilot applications will be initiated in 2003. In parallel, research and development program has been continued to support the regulatory implementation. A task force team has been formed and the official channel has also been opened for cooperation between regulatory body and utility-related organizations. Specifically, a couple of trial applications have been done up to now. First, the risk-based inspection (RBI) program has been implemented to improve regulatory inspection framework by utilizing risk information obtained from plant specific PSA and worldwide operating experiences. Next, optimization of technical specifications has been applied for surveillance test intervals (STIs) and allowed outage time (AOT). (author)

Software life after in-service

International Nuclear Information System (INIS)

Tseng, M.; Eng, P.

1993-01-01

Software engineers and designers tend to conclude a software project at the in-service milestone of the software life cycle. But the reality is that the 'life after in-service' is significantly longer than other phases of the life cycle, typically 20 years or more depending on the maintainability of the hardware platform and the designed life of the plant. During this period, the software asset (as with other physical assets in the plant) continues to be upgraded to correct deficiencies, meet new requirements, cope with obsolescence of equipment and so on. The software life cycle ends with a migration of the software to a different platform. It is typical in a software development project to put a great deal of emphasis on design methodologies, techniques, tools, development environment, standard procedures, and project management to ensure quality product is delivered on schedule and within budget. More often than not, a disproportion of emphasis is placed on the issues and needs of the in-service phase. Once the software is in-service, the designers move on to other projects, while the maintenance and support staff must manage the software. This paper examines the issues in three steps. First it presents a view of software from maintenance and support staff perspectives, including complexity of software, suitability of documentation, configuration management, training, difficulties and risks associated with making changes, required skills and knowledge. Second, it identifies the concerns raised from these viewpoints, including costs of maintaining the software, ability to meet additional requirements, availability of support tools, length of time required to engineer and install changes, and a strategy for the migration of software asset. Finally it discusses some approaches to deal with the concerns. (Author) 5 refs., fig

Dynamical systems probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ames, Arlo Leroy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-03-01

Probabilistic Risk Assessment (PRA) is the primary tool used to risk-inform nuclear power regulatory and licensing activities. Risk-informed regulations are intended to reduce inherent conservatism in regulatory metrics (e.g., allowable operating conditions and technical specifications) which are built into the regulatory framework by quantifying both the total risk profile as well as the change in the risk profile caused by an event or action (e.g., in-service inspection procedures or power uprates). Dynamical Systems (DS) analysis has been used to understand unintended time-dependent feedbacks in both industrial and organizational settings. In dynamical systems analysis, feedback loops can be characterized and studied as a function of time to describe the changes to the reliability of plant Structures, Systems and Components (SSCs). While DS has been used in many subject areas, some even within the PRA community, it has not been applied toward creating long-time horizon, dynamic PRAs (with time scales ranging between days and decades depending upon the analysis). Understanding slowly developing dynamic effects, such as wear-out, on SSC reliabilities may be instrumental in ensuring a safely and reliably operating nuclear fleet. Improving the estimation of a plant's continuously changing risk profile will allow for more meaningful risk insights, greater stakeholder confidence in risk insights, and increased operational flexibility.

Optical surface contouring for non-destructive inspection of turbomachinery

Science.gov (United States)

Modarress, Dariush; Schaack, David F.

1994-03-01

Detection of stress cracks and other surface defects during maintenance and in-service inspection of propulsion system components, including turbine blades and combustion compartments, is presently performed visually. There is a need for a non-contact, miniaturized, and fully fieldable instrument that may be used as an automated inspection tool for inspection of aircraft engines. During this SBIR Phase 1 program, the feasibility of a ruggedized optical probe for automatic and nondestructive inspection of complex shaped objects will be established. Through a careful analysis of the measurement requirements, geometrical and optical constraints, and consideration of issues such as manufacturability, compactness, simplicity, and cost, one or more conceptual optical designs will be developed. The proposed concept will be further developed and a prototype will be fabricated during Phase 2.

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

Risk-based Inspection Guide for the Susquehanna Station HPCI system

International Nuclear Information System (INIS)

Travis, R.; Higgins, J.; Gunther, W.; Shier, W.

1992-11-01

The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A system Risk-based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Susquehanna Steam Electric Station (SSES) which is operated by Pennsylvania Power ampersand Light (PP ampersand L). Included in this S-RIG is a discussion of the role of HPCI in mitigating accidents and a presentation of PRA-based failure modes which could prevent proper operation of the system. The S-RIG uses industry operating experience, including plant-specific illustrative examples, to augment the basic PRA failure modes. It is designed to be used as a reference for both routine inspections and the evaluation of the significance of component failures

Use of remote visual in-service inspection on nuclear power plants of the CEGB

International Nuclear Information System (INIS)

James, D.W.

1985-01-01

The main responsibility of the Remote Inspection Group is the design, development and procurement of the remote visual inspection equipment provided by the Generation Development and Construction Division as part of the extent of the supply for all the Central Electricity Generating Board's (CEGB) advanced gas-cooled reactors (AGR). The paper describes the operation of this equipment, together with the low light-level TV cameras that have been developed for carrying out routine remote visual inspections. The camera, known as the television remote inspection unit multi-purpose head (TRIUMPH), has been designed as a series of modules. With this system it is possible to take advantage of improvements in a particular part of the camera system and to arrange to backfit an improved module to existing TRIUMPHs. To minimize the time for carrying out routine inspections during shutdown, the AGRs have been provided with storage training and test facilities. These facilities are provided with full size mock-ups of the reactor internals so that the inspection equipment can be tested and the operating staff trained before the equipment is used on the reactor. One of the other responsibilities of the Remote Inspection Group is to carry out specific power plant remote visual inspections which are required to minimize costly plant shutdowns and construction delays. Examples are given of successful inspections that have been carried out. Over 12 years' experience has now been obtained in carrying out, at short notice, difficult inspections which involve tortuous access routes. The CEGB now holds a wide range of fibrescope and small TV cameras, together with the equipment for placing the viewing device in the correct location. A number of special fibrescopes have been developed for specific inspection needs and details of these, together with other fibrescopes owned by the CEGB, are provided. (author)

Super Phenix 1: In-service inspection of main and safety tanks weldments

Energy Technology Data Exchange (ETDEWEB)

Asty, M [DTech/STA, Centre d' Etudes Nucleaires de Saclay (France); Vertut, J [DPR/STEP, Centre d' Etudes Nucleaires de Saclay (France); Argous, J P [DRNR/STRS, Centre d' Etudes Nucleaires de Cadarache (France)

1980-11-01

In Service Inspection of the main tank of the Super Phenix 1 reactor is a new demand compared to Phenix: the authorities have asked that surface and internal defects be detected and their evolution monitored in the future. The presence of thermal baffles inside the main tank precludes the access on that side: the distance between the main and safety tanks takes into account the room needed for an In Service Inspection module. An inspection vehicle is presently under development, which includes ultrasonic examination (focussed probes) and visual examination (TV cameras) capabilities. We briefly describe the techniques that have been selected for ultrasonic testing and also for the vehicle and its guidance between the tanks. (author)

Super Phenix 1: In-service inspection of main and safety tanks weldments

International Nuclear Information System (INIS)

Asty, M.; Vertut, J.; Argous, J.P.

1980-01-01

In Service Inspection of the main tank of the Super Phenix 1 reactor is a new demand compared to Phenix: the authorities have asked that surface and internal defects be detected and their evolution monitored in the future. The presence of thermal baffles inside the main tank precludes the access on that side: the distance between the main and safety tanks takes into account the room needed for an In Service Inspection module. An inspection vehicle is presently under development, which includes ultrasonic examination (focussed probes) and visual examination (TV cameras) capabilities. We briefly describe the techniques that have been selected for ultrasonic testing and also for the vehicle and its guidance between the tanks. (author)

Application of risk-based inspection methods for cryogenic equipment

CERN Multimedia

CERN. Geneva

2016-01-01

Risk-based Inspection (RBI) is widely applied across the world as part of Pressure Equipment Integrity Management, especially in the oil and gas industry, to generally reduce costs compared with time-based approaches and assist in assigning resources to the most critical equipment. One of the challenges in RBI is to apply it for low temperature and cryogenic applications, as there are usually no degradation mechanisms by which to determine a suitable probability of failure in the overall risk assessment. However, the assumptions used for other degradation mechanisms can be adopted to determine, qualitatively and semi-quantitatively, a consequence of failure within the risk assessment. This can assist in providing a consistent basis for the assumptions used in ensuring adequate process safety barriers and determining suitable sizing of relief devices. This presentation will discuss risk-based inspection in the context of cryogenic safety, as well as present some of the considerations for the risk assessme...

Thermal Reactor Safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Thermal Reactor Safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

Structuring a risk-informed and performance-based process for optimization of regulation for Laguna Verde NPP

International Nuclear Information System (INIS)

Rodriguez-Hernandez, A.

2001-01-01

This work describes the plan for a process to incorporate into the regulatory activities the risk information derived from probabilistic risk assessments, as well as information generated by the periodic evaluation of the Maintenance Rule (MR, 10CFR50.65). The current status of the Laguna Verde NPP (LVNPP) risk analysis, PSA Level 1, allows determining in a reliable way the accident scenarios and the involved systems having significant impact on safety. The determination of system's risk significance allows carrying out a prioritization of safety issues to be evaluated and inspected; for example, operational events, changes to technical specifications, design modifications, inspection priorities, etc. In addition, complementary and basic information are the results generated by the performance monitoring of structures, systems and components (SSCs) under the scope of the MR. The SSCs performance trends are indicatives to focus evaluation and inspection activities on important issues. Then, with the reportability in short periods the performance evaluations of SSCs and the incorporation of a process of risk management, the evaluation and inspection activities will be directed to those risk significant systems showing degraded performance. Therefore, based on systems performance results and risk information, it is feasible to have certain flexibility or a better balance between the regulatory requirements. Inside this process, a consensus is needed with the utility to establish quality attributes for the plant-specific PSA, as well as the rules to be followed in the use of this tool and the kind of information to be reported for MR results. (author)

Performance-based inspection and maintenance strategies

International Nuclear Information System (INIS)

Vesely, W.E.

1995-01-01

Performance-based inspection and maintenance strategies utilize measures of equipment performance to help guide inspection and maintenance activities. A relevant measure of performance for safety system components is component unavailability. The component unavailability can also be input into a plant risk model such as a Probabilistic Risk Assessment (PRA) to determine the associated plant risk performance. Based on the present and projected unavailability performance, or the present and projected risk performance, the effectiveness of current maintenance activities can be evaluated and this information can be used to plan future maintenance activities. A significant amount of information other than downtimes or failure times is collected or can be collected when an inspection or maintenance is conducted which can be used to estimate the component unavailability. This information generally involves observations on the condition or state of the component or component piecepart. The information can be detailed such as the amount of corrosion buildup or can be general such as the general state of the component described as ' high degradation', ' moderate degradation', or ' low degradation'. Much of the information collected in maintenance logs is qualitative and fuzzy. As part of an NRC Research program on performance-based engineering modeling, approaches have been developed to apply Fuzzy Set Theory to information collected on the state of the component to determine the implied component or component piecepart unavailability. Demonstrations of the applications of Fuzzy Set Theory are presented utilizing information from plant maintenance logs. The demonstrations show the power of Fuzzy Set Theory in translating engineering information to reliability and risk implications

Reliability-based service life assessment of concrete structures in nuclear power plants: optimum inspection and repair

International Nuclear Information System (INIS)

Ellingwood, B.R.; Mori, Y.

1995-01-01

Research is being conducted to address aging management of safety-related reinforced concrete structures in nuclear power plants (NPPs). Documentation is being prepared to identify potential structural safety issues and to recommend criteria for use in evaluating reinforced concrete structures for continued service. Time-dependent reliability analysis provides the framework and quantitative tools for the condition assessment. The role of in-service inspection and repair in ensuring continued reliability in-service is examined. (author). 19 refs., 4 figs

Optimal Risk-Based Inspection Planning for Offshore Wind Turbines

DEFF Research Database (Denmark)

Rangel-Ramirez, Jose G.; Sørensen, John Dalsgaard

2008-01-01

, inspection and maintenance activities are developed. This paper considers aspects of inspection and maintenance planning of fatigue prone details in jacket and tripod types of wind turbine support structures. Based oil risk-based inspection planning methods used for oil & gas installations, a framework......Wind turbines for electricity production have increased significantly the last years both in production capability and size. This development is expected to continue also in the coining years. The Support structure for offshore wind turbines is typically a steel structure consisting of a tower...... for optimal inspection and maintenance planning of offshore wind turbines is presented. Special aspects for offshore wind turbines are considered: usually the wind loading are dominating the wave loading, wake effects in wind farms are important and the reliability level is typically significantly lower than...

The need to optimize inservice testing and inspection to enhance safety

International Nuclear Information System (INIS)

Perry, J.A.

1996-01-01

Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view

The need to optimize inservice testing and inspection to enhance safety

Energy Technology Data Exchange (ETDEWEB)

Perry, J.A.

1996-12-01

Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view.

Implementation of the Air Program Information Management System (APIMS) Inspection Module

Science.gov (United States)

2009-05-01

7 5 T H A I R B A S E W I N G Implementation of the Air Program Information Management System (APIMS) Inspection Module 2009 Environment...Implementation of the Air Program Information Management System (APIMS) Inspection Module 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER

Importance measures in risk-informed decision making: Ranking, optimisation and configuration control

Energy Technology Data Exchange (ETDEWEB)

Vaurio, Jussi K., E-mail: jussi.vaurio@pp1.inet.fi [Prometh Solutions, Hiihtaejaenkuja 3K, 06100 Porvoo (Finland)

2011-11-15

This paper describes roles, extensions and applications of importance measures of components and configurations for making risk-informed decisions relevant to system operations, maintenance and safety. Basic importance measures and their relationships are described for independent and mutually exclusive events and for groups of events associated with common cause failures. The roles of importances are described mainly in two groups of activities: (a) ranking safety significance of systems, structures, components and human actions for preventive safety assurance activities, and (b) making decisions about permissible permanent and temporary configurations and allowed configuration times for regulation, technical specifications and for on-line risk monitoring. Criticality importance and sums of criticalities turn out to be appropriate measures for ranking and optimization. Several advantages are pointed out and consistent ranking of pipe segments for in-service inspection is provided as an example. Risk increase factor and its generalization risk gain are most appropriately used to assess corrective priorities and acceptability of a situation when components are already failed or when planning to take one or more components out of service for maintenance. Precise definitions are introduced for multi-failure configurations and it is shown how they can be assessed under uncertainties, in particular when common cause failures or success states may be involved. A general weighted average method is compared to other candidate methods in benchmark cases. It is the preferable method for prediction when a momentary configuration is known or only partially known. Potential applications and optimization of allowed outage times are described. The results show how to generalize and apply various importance measures to ranking and optimization and how to manage configurations in uncertain multi-failure situations. - Highlights: > Rigorous methods developed for using importances

Importance measures in risk-informed decision making: Ranking, optimisation and configuration control

International Nuclear Information System (INIS)

Vaurio, Jussi K.

2011-01-01

This paper describes roles, extensions and applications of importance measures of components and configurations for making risk-informed decisions relevant to system operations, maintenance and safety. Basic importance measures and their relationships are described for independent and mutually exclusive events and for groups of events associated with common cause failures. The roles of importances are described mainly in two groups of activities: (a) ranking safety significance of systems, structures, components and human actions for preventive safety assurance activities, and (b) making decisions about permissible permanent and temporary configurations and allowed configuration times for regulation, technical specifications and for on-line risk monitoring. Criticality importance and sums of criticalities turn out to be appropriate measures for ranking and optimization. Several advantages are pointed out and consistent ranking of pipe segments for in-service inspection is provided as an example. Risk increase factor and its generalization risk gain are most appropriately used to assess corrective priorities and acceptability of a situation when components are already failed or when planning to take one or more components out of service for maintenance. Precise definitions are introduced for multi-failure configurations and it is shown how they can be assessed under uncertainties, in particular when common cause failures or success states may be involved. A general weighted average method is compared to other candidate methods in benchmark cases. It is the preferable method for prediction when a momentary configuration is known or only partially known. Potential applications and optimization of allowed outage times are described. The results show how to generalize and apply various importance measures to ranking and optimization and how to manage configurations in uncertain multi-failure situations. - Highlights: → Rigorous methods developed for using importances

Auxiliary feedwater system risk-based inspection guide for the Diablo Canyon Unit 1 Nuclear Power Plant

International Nuclear Information System (INIS)

Gore, B.F.; Vo, T.V.; Harrison, D.G.

1990-08-01

This document presents a compilation of auxiliary feedwater (AFW) system failure information which has been screened for risk significance in terms of failure frequency and degradation of system performance. It is a risk-prioritized listing of failure events and their causes that are significant enough to warrant consideration in inspection planning at Diablo Canyon. This information is presented to provide inspectors with increased resources for inspection planning at Diablo Canyon. The risk importance of various component failure modes was identified by analysis of the results of probabilistic risk assessments (PRAs) for many pressurized water reactors (PWRs). However, the component failure categories identified in PRAs are rather broad, because the failure data used in the PRAs is an aggregate of many individual failures having a variety of root causes. In order to help inspectors to focus on specific aspects of component operation, maintenance and design which might cause these failures, an extensive review of component failure information was performed to identify and rank the root causes of these component failures. Both Diablo Canyon and industry-wide failure information was analyzed. Failure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorized as common cause failures, human errors, design problems, or component failures. This information permits an inspector to concentrate on components important to the prevention of core damage. Other components which perform essential functions, but which are not included because of high reliability or redundancy, must also be addressed to ensure that degradation does not increase their failure probabilities, and hence their risk importances. 23 refs., 1 fig., 1 tab

Mobile X-ray inspection of light weight materials

International Nuclear Information System (INIS)

Ewert, Uwe; Redmer, Bernhard; Raedel, Christoph; Osterloh, Kurt; Schnars, Ulf; Henrich, Rudolf; Schimmelmann, Olaf; Bavendiek, Klaus; Jahn, Mirko

2008-01-01

Digital detectors such as phosphor imaging plates (IP) and digital detector arrays (DDA) allow radiographic inspection with higher efficiency and improved image quality in comparison to the classic film technique. Mobile X-ray flash tubes are used routinely for veterinarian and security applications. New high sensitive IPs and DDAs enable to apply them for inspection of light materials with low X-ray attenuation as in aluminium, plastics and composites. A versatile computed tomography (CT) system was developed for in situ inspection of large aircraft components under production conditions. A gate based planar computed tomograph was developed and tested for inspection of integrity of the stringer incorporation. Successful test trials were performed to prove the detection rate of cracks in embedded stringers. Honey comb structures of aircrafts have to be inspected for water inclusions during in-service inspections. Thermography is a powerful method for in house inspections when variations in temperature caused e.g. by sunshine can be excluded. A new X-ray diaphragm was developed for mobile back scatter measurements of large components. This method is insensitive to heat alterations in the field and thus can be applied also outdoors. (orig.)

In-service inspection of condenser tubes by means of electrochemical methods

International Nuclear Information System (INIS)

Taelemans, G.

The commissioning of an increasing number of large nuclear power plants involves an increased significance of such condenser tube problems as: - erosion on tube ends, - generalized corrosion and pitting corrosion, - deposits in the tubes. In order to solve such problems, investigations were performed especially focused on a measurement technique that enables in-service behaviour of condenser tubes to be monitored. For such a purpose, measurement of the polarization resistance has been adopted. The existing corrosion products and scaled-off iron oxides were eliminated by means of a carborundum balls processing, as clearly appears from polarization resistance reduction. Then iron sulphate was injected in order to build a new and better protective layer. In addition, the tube was kept clean by means of foam rubber balls. There is a second implementation area: fouled condenser tubes. A significant polarization resistance reduction is noted during the acid cleaning. (orig.) [de