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Sample records for risk-informed inservice inspection

  1. Risk informed In-service Inspection

    International Nuclear Information System (INIS)

    Corak, Z.

    2003-01-01

    insights. It is expected that licensees will ask for modifications of plant's design, operation, and other activities that require NRC approval to incorporate risk insights into their ISI programs. In-service inspection program in which is incorporated risk insights is known as risk-informed in-service inspection programs, RI-ISI. Until the RI-ISI is approved for generic use, the staff expects that licensees will ask for modifications of ISI programs by requesting NRC to approve alternative inspection programs. Alternative inspection programs must meet 10 CFR 50.55a (a)(3)(i) which says that 'the applicant shall demonstrate that the proposed alternatives would provide an acceptable level of quality and safety'. As always, licensees should identify how the chosen approach, methods, data, and criteria are appropriate for the decisions that have to be made. RI-ISI can provide reduction of scope and frequency of inspections that may lead to higher risk. The higher risk means that a probability of failure or consequences are higher and shutdown of nuclear power plants is more likely to occur. RI-ISI can also provide extensive scope of inspection or increased frequency of inspection. (author)

  2. Risk informed in-service inspection and testing in Spain

    International Nuclear Information System (INIS)

    Bros, Juan; Marcelles, Ignacio

    2002-01-01

    The Spanish nuclear regulatory authority, the Consejo de Seguridad Nuclear (CSN), requires the use of codes and standards in force in the country of origin of the plant technology. For this reason, the in-service inspection and testing programs applied at Spanish nuclear power plants basically adhere to the requirements of the ASME XI and ASME OM Codes. It is not surprising that when the earliest developments aimed at drawing up risk informed inservice inspection and testing programs were initiated within the framework of ASME, the Spanish industry should follows such developments very closely. In fact, persons within the Spanish nuclear industry joined different ASME committees involved in the development and approval of the various code cases encompassing these developments. Developments specific to the Spanish nuclear power plants were initiated at a time when the aforementioned reference documentation was in a very advanced stage of development/approval. Two clearly differentiated lines of work got under way: On the one hand, and as regards risk informed in-service testing programs, the American standards were used as the sole reference. In the case of risk informed in-service inspection programs, the Spanish nuclear power plant-owning utilities and the Consejo de Seguridad Nuclear decided to draw up a Spanish guideline that, although using the ASME developments as a reference, would have its own specific characteristics. In relation to the above, and referring to the chronology of the events, the activities performed to date in Spain are described

  3. Study on practical application of risk informed inservice inspection

    International Nuclear Information System (INIS)

    Sato, Chikahiro; Machida, Hideo; Takeda, Shuhei; Miyata, Koichi; Nishino, Shoichiro

    2009-01-01

    This paper describes pilot study relevant to the application of risk informed inservice inspection (RI-ISI) to class 1 piping systems in a Japanese typical BWR5 plant. The benefits of making use of risk information are improvement in plant safety, quality of inspection and explanation of security activities in nuclear power plants. The current RI-ISI procedures and rules were developed to take advantage of lessons learned from PSA data and piping failure experiences, and are expected to rationalize of security activities for plant operation and maintenance. To introduce RI-ISI, it is necessary to collaborate with industry, academia and government. Development of the technical basis is one of the key issues to become practical of RI-ISI programs. (author)

  4. Risk informed approach to the in-service inspection activities

    International Nuclear Information System (INIS)

    Korosec, D.; Vojvodic Tuma, J.

    2004-01-01

    In the present paper, the aspects of Risk Informed In-Service Inspection (RI-ISI) are discussed. Slovenian Nuclear Safety Administration (SNSA) and its authorized organization for the ISI activities, Institute of Metals and Technologies (IMT), 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 existing practice in nuclear world, the activities are started to asses the piping of systems in the light of probability of failure. This is so called Risk Informed approach. By the design established criteria, standards and practice gives good fundaments 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 this deterministic approach, the principles of risk evaluation methods should be used. NPP Krsko has, as practically majority of NPP's, probabilistic risk assessment (PRA) studies for all safety important systems and components. The methods and results from these studies can be efficiently used to upgrade classical deterministic results, based on which the in-service program as a whole is usually done. In addition to the above mentioned, risk assessment and evaluation of the piping shall be done, which is not covered by the existing PRA analysis. To do this it is necessary to made risk evaluation of the piping segments, based on previous structural element probability assessment. Probabilistic risk assessment is important

  5. Spanish guideline for the definition and assessment of risk-informed inservice inspection programs for piping

    Energy Technology Data Exchange (ETDEWEB)

    Figueras, J.M.; Olivar, F.; Mendoza, C.; Vazquez, T. [CSN, Madrid (Spain); Morales, L. [UNESA, Madrid (Spain); Bros, J.; Mach, I. [Tecnatom, Madrid (Spain); Gutierrez, E. [Iberinco, Madrid (Spain)

    2001-07-01

    The recent emphasis on risk-informed in-service inspection has been taken on by the Spanish Utilities and the Spanish Regulatory Body in their proposal for a co-operation project for defining a Risk-Informed In-Service Inspection Guideline for Piping. This paper describes first the main features of this Spanish guideline and then the results of the pilot applications developed in order to check the consistency of the guideline. (authors)

  6. Spanish guideline for the definition and assessment of risk-informed inservice inspection programs for piping

    International Nuclear Information System (INIS)

    Figueras, J.M.; Olivar, F.; Mendoza, C.; Vazquez, T.; Morales, L.; Bros, J.; Mach, I.; Gutierrez, E.

    2001-01-01

    The recent emphasis on risk-informed in-service inspection has been taken on by the Spanish Utilities and the Spanish Regulatory Body in their proposal for a co-operation project for defining a Risk-Informed In-Service Inspection Guideline for Piping. This paper describes first the main features of this Spanish guideline and then the results of the pilot applications developed in order to check the consistency of the guideline. (authors)

  7. A risk-informed approach to optimising in-service inspection of piping

    International Nuclear Information System (INIS)

    Billington, A.; Monette, P.

    1999-01-01

    Traditional criteria for the selection of in-service inspection locations in piping, have come to be regarded as being out-of-touch with current knowledge of piping failures and with current measures of safety importance. An alternative , risk-informed, method has been developed and successfully licensed, that systematically establishes an inspection plan addressing all safety-related piping systems, in a way that is optimized with respect to the safety gain achieved through in-service inspection. The principles of the method are discussed and the results of several applications are summarized, all of which demonstrate that the risk-informed program would lead to significant improvements in the overall level of plant safety, while at the same time re-distributing the inspections in such a way that reduces both plant costs and radiation exposure to personnel.(author)

  8. Risk-informed in-service inspections of nuclear power plants: European activities

    International Nuclear Information System (INIS)

    Simola, K.; Gandossi, L.

    2010-01-01

    Risk-informed in-service inspection (RI-ISI) methods aim at enhancing ISI effectiveness by taking into account the risk importance of possible inspection sites. These methods are widely applied in US, but in Europe the situation is different, as there are many regulatory environments implying a variety of ISI codes and standards and national guidelines. The objective of the European Network for Inspection Qualification, ENIQ, is to co-ordinate and to manage at European level expertise and resources for the qualification of non-destructive inspection techniques and procedures primarily for the in-service inspection of nuclear components. ENIQ has established a Task Group on Risk (TGR) to work towards developing best practice for RI-ISI methodologies. TGR has published a European framework document for risk-informed in-service inspection, and the group has been working at producing more detailed recommended practices and discussion documents on several RI-ISI related issues. In addition, TGR has been active in initiating international projects, such as the JRC-OECD/NEA coordinated RI-ISI benchmark exercise (RISMET). This paper describes the activities and publications of TGR to date, and summarises the contents and main results of the RISMET RI-ISI benchmark exercise. (orig.)

  9. Increase plant safety and reduce cost by implementing risk-informed in-service inspection programs

    International Nuclear Information System (INIS)

    Billington, A.; Monette, P.

    2001-01-01

    The idea behind the program is that it is possible to 'inspect less, but inspect better'. In other words, the risk-informed In-Service Inspection (ISI) process is used to improve the effectiveness of examination of piping components, i.e. concentrate inspection resources and enhance inspection strategies on high safety significant locations, and reduce inspection requirements on others. The Westinghouse Owners Group (WOG) risk-informed ISI process has already been applied for full scope (Millstone 3, Surry 1) and limited scope (Beznau, Ringhals 4, Asco, Turkey Point 3). By examining the high safety significant piping segments for the different fluid piping systems, the total piping core damage frequency is reduced. In addition, more than 80% of the risk associated with potential pressure boundary failures is addressed with the WOG risk-informed ISI process, while typically less that 50% of this same risk is addressed by the current inspection programs. The risk-informed ISI processes are used to improve the effectiveness of inspecting safety-significant piping components, to reduce inspection requirements on other piping components, to evaluate improvements to plant availability and enhanced safety measures, including reduction of personnel radiation exposure, and to reduce overall Operation and Maintenance (O and M) costs while maintaining regulatory compliance. A description of the process as well as benefits from past projects is presented, since the methodology is applicable for WWER plant design. (author)

  10. Increase plant safety and reduce cost by implementing risk-informed In-Service Inspection programs

    International Nuclear Information System (INIS)

    Billington, A.; Monette, P.; Doumont, C.

    2000-01-01

    The idea behind the program is that it is possible to 'inspect less, but inspect better'. In other words, the risk-informed In-Service Inspection (ISI) process is used to improve the effectiveness of examination of piping components, i.e. concentrate inspection resources and enhance inspection strategies on high safety significant locations, and reduce inspection requirements on others. The Westinghouse Owners Group (WOG) risk-informed ISI process has already been applied for full scope (Millstone 3, Surry 1) and limited scope (Beznau, Ringhals 4, Asco, Turkey Point 3). By examining the high safety significant piping segments for the different fluid piping systems, the total piping core damage frequency is reduced. In addition, more than 80% of the risk associated with potential pressure boundary failures is addressed with the WOG risk-informed ISI process, while typically less than 50% of this same risk is addressed by the current inspection programs. The risk-informed ISI processes are used: to improve the effectiveness of inspecting safety-significant piping components; to reduce inspection requirements on other piping components; to evaluate improvements to plant availability and enhanced safety measures, including reduction of personnel radiation exposure; and to reduce overall Operation and Maintenance (O and M) costs while maintaining regulatory compliance. A description of the process as well as benefits of past projects is presented, since the methodology is applicable for VVER plant design. (author)

  11. Risk informed inservice inspection evaluation of the ComEd nuclear plants

    International Nuclear Information System (INIS)

    Fleming, K.N.; Hutchinson, J.J.

    2000-01-01

    In November, 1999 Commonwealth Edison Company began the largest scope project yet performed to implement risk informed in-service inspection programs for piping systems. A contractor team led by ERIN Engineering and Research was selected to perform full scope risk informed evaluations on ten reactor units at five sites including 3 BWR and 2 PWR stations. This risk informed evaluation has applied a number of advancements to RISI technology in key areas including risk quantification of inspection program changes, element selection, treatment of multiple damage mechanisms at a given location, and in streamlining an approach to RISI initially developed by EPRI. The scope of the evaluation covered more than 18,000 ASME Class 1 and 2 welds not counting those that will continue to be inspected as part of augmented programs for flow accelerated corrosion, stress corrosion cracking, and service water systems. When the current ASME Section XI based exams are replaced by the risk informed program, while some new inspection locations will be added, substantial cost savings and person-rem exposures are projected from the elimination of most of the current exams. ComEd expects to recover its investment in the RISI evaluation for each of the 10 reactor units in 1 to 2 refueling outages, from the cost savings of reduced inspections. At the same time, the risk of a severe accident due to pipe ruptures is not expected to exhibit significant changes. (author)

  12. International developments on implementation of Wog risk-informed inservice inspection methodology

    International Nuclear Information System (INIS)

    Balkey, K.R.; Bishop, B.A.; Canton, M.A.; Closky, N.B.; Haessler, R.L.; Kolonay, J.F.; Sharp, G.L.; Stevenson, P.R.

    2001-01-01

    The full text follows. The Westinghouse Owners Group (WOG) risk-informed inservice inspection (ISI) methodology was granted approval by the U.S. Nuclear Regulatory Commission in 1998 thereby providing an alternative to ASME Section XI Code requirements for the selection of examination locations in nuclear plant piping systems. This paper builds upon a technical paper presented at ICONE-8 that reported on the first wave of risk-informed ISI applications under development primarily focusing on those underway within the U.S. Since that time, many applications have continued within the U.S., however, much progress has been made in applying the WOG risk-informed ISI approach in several other countries. While a summary of results across the various applications will be provided, the paper will focus on the development and implementation of the WOG risk-informed ISI methodology across Europe and in Asia for both full scope and limited Class 1 scope applications. An update on future risk-informed applications, such as modifying requirements for augmented examinations for high energy line break exclusion regions and in risk-informing the safety classification of pressure boundary components in support of risk-informed regulation initiatives, will also be provided. (authors)

  13. 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)

  14. 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)

  15. Application of risk-informed methods to in-service piping inspection in Framatome type nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Jin Hoi; Lee, Jeong Seok; Yun, Eun Sub

    2014-01-01

    The Pressurized water reactor owners group (PWROG) developed and applied a risk-informed in-service inspection (RI-ISI) program, as an alternative to the existing ASME Section XI sampling inspection method. The RI-ISI programs enhance overall safety by focusing inspections of piping at high safety significance (HSS) locations where failure mechanisms are likely to be present. Additionally, the RI-ISI program can reduce nondestructive evaluation (NDE) exams, man-rem exposure for inspectors, and inspection time, among other benefits. The RI-ISI method of in-service piping inspection was applied to 3 units (KSNPs: Korea standard nuclear power plants) and is being deployed to the other units. In this paper, the results of RI-ISI for a Framatome type (France CPI) nuclear power plant are presented. It was concluded that application of RI-ISI to the plant could enhance and maintain plant safety, as well as provide the benefits of greater reliability.

  16. RIBA Project - Risk-Informed approach for In-Service Inspection of Nuclear Power Plant Components. Project summary

    International Nuclear Information System (INIS)

    Lidbury, D.; Smith, G.

    2001-12-01

    The need for a European review of a Risk-Informed Approach for In-Service Inspection of Nuclear Power Plant Components (RIBA) was identified in 1998. This was as a priority item in the programme of activities conducted in the framework of the Council Resolutions of 22 July 1975 and of 18 June 1992 on the Technological Problems of Nuclear Safety. The RIBA Project was established in November 1999 as a 24-month Study Contract funded by the European Commission within the frame of the former DG XI WGCS (Working Group on Codes and Standards). The Study Contract was subsequently managed for the EC by DG TREN. The participants in RIBA were Serco Assurance (project coordinator), Ringhals AB, EDF, Tecnatom SA and Westinghouse Electric Europe. The work is presented in a summary report with the detailed results contained in three companion reports as follows: main conclusions and recommendations, Review of Existing Risk-Informed Methodologies, A Comparative Study of Risk-Informed In-Service Inspection Applications, Conclusions and Recommendations for Risk-Informed in-service inspection methodology applied to Nuclear Power Plants in Europe. (author)

  17. Report on the Regulatory Experience of Risk-Informed In-service Inspection of Nuclear Power Plant Components and Common Views (consensus document)

    International Nuclear Information System (INIS)

    2004-08-01

    The present report represents the work product of the activities conducted by the Task Force. The TF performed a review and inventory of the existing approaches to risk-informed inservice inspection and testing, and completed its work in 1999 with a Current Practices Document 2, titled Report on risk-informed in-service inspection and in-service testing (EUR 19153 EN). In November 2001, the NRWG held a Special session on risk-informed applications, with emphasis on risk-informed inservice inspection, where results and experiences from pilot studies on risk-informed inservice inspection (RI-ISI), performed in several European countries, were presented and discussed. As a follow-up in May 2002, the TF was reconvened with the objectives to analyse from the regulatory point of view key aspects associated with the application of risk-informed inservice inspection, and to go beyond a state of the art report, presenting a series of recommendations of good practices or common positions reached by the regulators represented in the Task Force. (author)

  18. Application of risk-informed in-service inspection approach. Pilot study on low pressure emergency core cooling system of NPP Temelin

    International Nuclear Information System (INIS)

    Horacek, L.; Palyza, J.; Zdarek, J.; Vizina, M.

    2004-01-01

    In-service inspection (ISI) programme of piping systems based on Riskinformed In-service Inspections (RI-ISI) approach represents for the WWER NPP licensee first overall systematic conceptual solution of the problems associated with design and design modifications of ISI programme for piping systems. The approach enables to consider possible savings and comparison of in-service inspection programmes according to more objective criteria including those calculated in PSA, in contrast to present much empiric experience collected by manufacturers, operators and ISI vendors in the past. Service experience has shown limited correlation between the deterministic ISI requirements and actual field failures or degradation mechanisms. Where field failures have been observed in piping, they have generally been due to either material concerns (e.g., Intergranular Stress Corrosion Cracking) or stress/cycling mechanisms not identified in the original design basis documents (e.g., thermal stratification), and therefore would not be selected for inspection under current requirements. Risk-informed in-service inspections represent integrated multidisciplinary approach to the in-service inspection programme fulfilling the enhanced requirements of the Czech Regulatory Body (SUJB) for defence in depth concept applied to weld joints of safety related components, application of qualified NDT methods/techniques and introducing of inspections for cause approach. Inspections for cause take into account analysis of both active and potential degradation mechanisms specific to piping system concerned. Simultaneously, this approach enables the licensee to reach, in cases justified from safety point of view, economic savings resulting from lowering of both number and extent of performed in-service inspections and optimisation of their performance. RI ISI pilot study based in majority on EPRI methodology implemented for WWER type reactors has been performed for LP ECCS (Low Pressure Emergency

  19. 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

  20. 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.)

  1. 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)

  2. 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

  3. 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

  4. 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.

  5. The applications of risk-informed in-service-inspection

    International Nuclear Information System (INIS)

    Ting, K.; Ko, T.-H.; Li, Y.-C.; Wu, W.-F.; Lu, Y.-L.; Chien, F.-T.

    2005-01-01

    The US NRC and nuclear industry encouraged the applications of risk-informed In-Service Inspection (RI-ISI) which can be an alternative program to the ASME Section XI In-service Inspection requirements. The Implementation of RI-ISI can improve the substantial cost as well as does reductions. From the aspect of defense in depth for nuclear safety, the achievements of these procedures can identify the inspection rank to promote the integrity of the current inspection program. Thus, this study utilizes this techniques to implement risk assessment on safety class 1 recirculation piping welds where sensitized to IGSCC of Taibei BWR-6 nuclear power plant. In the evaluation process, WinPraise code is used to calculate the failure probabilities of all welds. The result of risk evaluations can be referred to the further regulatory and plant operation. (authors)

  6. 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)

  7. 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)

  8. 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

  9. 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.)

  10. 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)

  11. 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)

  12. Targeted In-service Inspections Using Risk Insights

    International Nuclear Information System (INIS)

    Kulat, S.; Montgomery, B.; Robin Graybeal, M.

    2012-01-01

    This paper includes a discussion of the historical background and rationale for development of a targeted In-service Inspection (ISI) program using risk insights, known as Risk-Informed In-service Inspection (RI-ISI). RI-ISI programs are optimized inspection programs which target specific welds for inspection based upon potential degradation mechanisms and consequences of failure. Inspections are performed on those welds that are the highest contributors to plant risk. Additionally the inspections are tailored to detect the specific postulated degradation mechanisms. As a result, the numbers of inspections are reduced along with the associated cost and radiation exposure, while maintaining or improving the level of quality and safety. Provided in this paper are the basic principles of RI-ISI program development, and a summary of the impact of the implementation of such programs. For example, implementation of a Risk-Informed In-service Inspection program results in a reduction of both cost and radiation exposure. Cost savings are estimated at between USD 1,000,000 and USD 2,300,000 per unit per ten year interval for a Class 1 and2 RI-ISI application. Cost savings are estimated based on average cost per weld of USD 7600 for examination, including the following activities: erection and removal of scaffolding, removal and replacement of insulation, removal and replacement of interferences, weld preparation, examination, documentation, craft support. Reduction in radiation exposure is estimated at 75% to 90% for a Class 1 and2 RI-ISI application. Reduction in radiation exposure is due to the following factors: number of welds selected for examination decreases by 60% to 75%, surface examinations essentially eliminated, within a given risk category, welds can be selected for examination based on additional factors such as the minimization of radiation exposure.(author).

  13. 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.

  14. 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

  15. 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.

  16. Improvement and optimization for in-service inspection of M310 nuclear power station

    International Nuclear Information System (INIS)

    Wang Chen; Sun Haitao; Gao Chen; Deng Dong

    2015-01-01

    In-service inspection (ISI) is an important method to ensure the safety of the mechanical equipment in nuclear power stations. According to the in-service inspection experience feedback from the domestic nuclear power stations, the reasonableness of some provisions in the RSE-M code are discussed and the applications of risk-informed in-service inspection (RI-ISI) are introduced, and the advices for the optimization of the ISI of the domestic M310 nuclear power stations are proposed. (authors)

  17. 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

  18. 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)

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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.)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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.)

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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.

  15. 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.)

  16. 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.)

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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.

  2. 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...

  3. 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...

  4. 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

  5. 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.)

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. 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)

  8. 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.)

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. 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)

  3. 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

  4. Experience on implementing risk-oriented in-service inspection at NPP in USA

    International Nuclear Information System (INIS)

    Grebenyuk, Yu.P.; Zaritskij, N.S.; Kovyrshin, V.G.; Kostenko, S.P.

    2001-01-01

    Experience of implementing risk-oriented in-service inspection (RIISI) of pipelines at the US NPP analyzed. RIISI processes according to the methodologies developed by Westinghouse and EPRI are described and compared. The information is addressed concerning practical matters of RIISI implementing at NPP Beaver Valley and Arkansas Nuclear One. Regulatory activity of the US NRC while implementing NRC at the US NPP. It is shown that using RIISI at NPP is aimed at improvement of examination efficiency, decreasing costs and occupational dose loads while inspecting pipelines

  5. 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

  6. 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.)

  7. 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.

  8. 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

  9. 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.)

  10. 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.)

  11. 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)

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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.)

  4. 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)

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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.)

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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)

  11. 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

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

    International Nuclear Information System (INIS)

    2003-08-01

    In November 1992, the Nuclear Regulators Working Group (NRWG) decided to set up a task force on qualification of non-destructive testing (NDT) systems for pre and in-service inspection of light water reactors. The first task was to agree on the philosophy and principles governing the qualification of techniques, equipment, software, procedures, and personnel for NDT to be used for the inspection of structural components that are important to safety in nuclear power plants; and to establish a common view on essential aspects of NDT qualifications. The first task, which also included a comparison of the common views of the European regulators with the qualification approach outlined in Appendix VIII to Section XI of the ASME Code, was completed in 1996. The result was published in the report ''Common position of European regulators on qualification of NDT systems for pre- and in-service inspection of light water reactor components''2. In parallel, the European nuclear power industries had set up a working group, the European Network for Inspection Qualification (ENIQ), to discuss and agree on how to perform inspection qualifications. In 1995, ENIQ finalized its first version of ''European methodology for qualification of non-destructive tests''3. A second version 4 was then published in 1997. This second version is in relatively close agreement with the principles given in the regulators common position document. With these two basic documents, a platform was established for the further development of qualification strategies in the European countries. The second task of the NRWG Task Force was to follow and evaluate the first ENIQ pilot study from a regulatory point of view. The objective of this pilot study was to explore ways of how to apply the European qualification methodology and to test its feasibility. The pilot study commenced late 1996 and was planned to be finalized a year later. Depending on unforeseen difficulties, the pilot study has been delayed

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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%)

  19. 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

  20. 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

  1. 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.)

  2. 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)

  3. 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.)

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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.).

  12. 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)

  13. 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

  14. 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)

  15. 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.

  16. 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

  17. 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

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

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

    International Nuclear Information System (INIS)

    Kojima, Y.; Matsuyama, A.

    1995-01-01

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

  13. 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

  14. 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)

  15. 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)

  16. Summary report for 1990 inservice inspection (ISI) of SRS 100-K reactor tank

    International Nuclear Information System (INIS)

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

    1990-01-01

    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 purpose of this inspection was to determine if selected welds in the K Reactor tank wall contained any indications of IGSCC. These portions 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. This inspection comprised approximately 60% of the accessible weld length in the K Reactor tank. Initial setup of the tank, which prior to inspection contained Mark 60B target assemblies but no Mark 22 fuel assemblies, began on January 14, 1990. The inspection was completed on March 9, 1990

  17. Ageing Management Programme: An Experience of In-Service Inspection of the Kartini Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Nitiswati, S., E-mail: nitis@batan.go.id [Centre for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (BATAN), Jakarta (Indonesia); Syarip,; Tjiptono, T.; Wantana, [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

    2014-08-15

    This paper discusses an experience on ISI of the Kartini reactor heat exchanger, as part of the implementation of an ageing management programme. Kartini reactor is located in Yogyakarta, Indonesia. The heat exchanger was constructed for 250 kW capacity. The type of heat exchanger is shell with tube recirculation. Tube material is stainless steel 304, and the shell and baffle plate materials are carbon steel. The heat exchanger has 72 tubes, its outer and inner diameters respectively are 19 and 16 mm, and tube thickness is 1.5 mm. The aim of ISI was to obtain and evaluate the heat exchanger’s condition including the detection of any possible local tube thinning, pitting corrosion or gradual thinning, and determine whether any degradation or deterioration of the heat exchanger could have a significant impact to safety. The heat exchanger was inspected utilizing eddy current equipment in 2003 and 2006. Inspection results in 2003 determined that 12 heat exchanger tubes have a thinning degradation level ranging from 10% up to 60% of the outer diameter due to pitting corrosion. Deterioration of baffle plates has been linked to general corrosion attack. Inspection results in 2006 showed a consistent thinning degradation level with the previous inspection in 2003. So far heat exchanger performance is still satisfactory, as defined by the transfer of primary heat for a 2°C difference between inlet and outlet as required. (author)

  18. 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.

  19. 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)

  20. 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

  1. In-service inspection of the vessels of nuclear submarine reactors

    International Nuclear Information System (INIS)

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

    1977-02-01

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

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

    International Nuclear Information System (INIS)

    Viertl, J.R.M.

    1988-01-01

    The most commonly applied inspection method for ferromagnetic turbine and generator rotor bores is the magnetic particle test technique. This method is subjective, depends on the test operator's skill and diligence in identifying test indications, and suffers from poor repeatability, especially for small indications. Automation would improve repeatability. However, magnetic particle tests are not easily automated, because the data are in the form of sketches, photographs, and written and oral descriptions of the indications. Eddy current inspection has obvious potential to replace magnetic particle methods in this application. Eddy current tests can be readily automated, as the data are in the form of voltages that can be recorded, digitized, and manipulated by a computer. The current project continues the investigation of the correlation between eddy current and magnetic particle inspection. Two systems have been combined to acquire eddy current data automatically. This combination of systems consists of the Nortec-25L Eddyscope (to provide the analog eddy current signals) and the General Electric DATAQ (TM) System (to perform the automatic data acquisition). The automation of the system is discussed

  3. 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.

  4. 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)

  5. 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.

  6. French codes and standards for design, construction and in-service inspection of nuclear power plants

    International Nuclear Information System (INIS)

    Hugot, G.; Grandemange, J. M.

    1995-01-01

    In 1970, France decided that its future power plants would be of the Pressurized Water Reactor type. This choice proved to be successful since it resulted in more than 60 PWR units in operation or under construction in France and abroad. At the beginning of such a program, the French engineering and manufacturing industry, the national electrical utility and the Safety Authorities had to face the many challenges imposed by the implementation of an imported technology. The government reorganised the licensing process. FRAMATOME, the NSSS vendor, and EDF (Electricite de France), the national utility, decided to create 'AFCEN', the French Association for Design and Construction Rules for Nuclear Island Components. These rules, the RCC's (Regles de Construction et de conception), which are approved by French Safety Authorities deal with mechanical and electrical equipment as well as with nuclear fuel and civil works. They are now being supplemented by in service inspection rules, the RSE's (Regles d'inspection en Service). The paper presents these Codes and their main updating following experience of application, technical progress and evolution of standards. Status of discussion concerning reference to European standardisation and developments of rules applicable to the EPR project will also be discussed

  7. Evaluation of sampling plans for in-service inspection of steam generator tubes

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Heasler, P.G.; Baird, D.B.

    1994-02-01

    This report summarizes the results of three previous studies to evaluate and compare the effectiveness of sampling plans for steam generator tube inspections. An analytical evaluation and Monte Carlo simulation techniques were the methods used to evaluate sampling plan performance. To test the performance of candidate sampling plans under a variety of conditions, ranges of inspection system reliability were considered along with different distributions of tube degradation. Results from the eddy current reliability studies performed with the retired-from-service Surry 2A steam generator were utilized to guide the selection of appropriate probability of detection and flaw sizing models for use in the analysis. Different distributions of tube degradation were selected to span the range of conditions that might exist in operating steam generators. The principal means of evaluating sampling performance was to determine the effectiveness of the sampling plan for detecting and plugging defective tubes. A summary of key results from the eddy current reliability studies is presented. The analytical and Monte Carlo simulation analyses are discussed along with a synopsis of key results and conclusions

  8. 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

  9. 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

  10. 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.

  11. PAT portable appliance testing : in-service inspection and testing of electrical equipment

    CERN Document Server

    Scaddan, Brian

    2015-01-01

    The Electricity at Work Regulations 1989 requires any electrical system to be constructed, maintained and used in such a manner as to prevent danger. This means that inspection and testing of systems, including portable appliances, is needed in order to determine if maintenance is required. This book explains in clear language what needs to be done and includes expert advice on legislation as well as actual testing. The book contains an appendix providing the electrical fundamentals needed by non-specialists and also has sample questions (with answers) for the C&G 2377 exam. It is also an ideal revision guide for the non-specialist, such as maintenance staff and caretakers who carry out these tasks part-time, alongside their many other duties.

  12. Evaluation of sampling schemes for in-service inspection of steam generator tubing

    International Nuclear Information System (INIS)

    Hanlen, R.C.

    1990-03-01

    This report is a follow-on of work initially sponsored by the US Nuclear Regulatory Commission (Bowen et al. 1989). The work presented here is funded by EPRI and is jointly sponsored by the Electric Power Research Institute (EPRI) and the US Nuclear Regulatory Commission (NRC). The goal of this research was to evaluate fourteen sampling schemes or plans. The main criterion used for evaluating plan performance was the effectiveness for sampling, detecting and plugging defective tubes. The performance criterion was evaluated across several choices of distributions of degraded/defective tubes, probability of detection (POD) curves and eddy-current sizing models. Conclusions from this study are dependent upon the tube defect distributions, sample size, and expansion rules considered. As degraded/defective tubes form ''clusters'' (i.e., maps 6A, 8A and 13A), the smaller sample sizes provide a capability of detecting and sizing defective tubes that approaches 100% inspection. When there is little or no clustering (i.e., maps 1A, 20 and 21), sample efficiency is approximately equal to the initial sample size taken. Thee is an indication (though not statistically significant) that the systematic sampling plans are better than the random sampling plans for equivalent initial sample size. There was no indication of an effect due to modifying the threshold value for the second stage expansion. The lack of an indication is likely due to the specific tube flaw sizes considered for the six tube maps. 1 ref., 11 figs., 19 tabs

  13. 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.

  14. 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.)

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

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

    International Nuclear Information System (INIS)

    Rapin, Michel; Saglio, Robert.

    1978-01-01

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

  3. 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

  4. Role of in-service inspection and maintenance in reliability assurance of concrete structures in nuclear plants

    International Nuclear Information System (INIS)

    Ellingwood, B.R.; Mori, Y.

    1993-01-01

    The degradation in strength of a component with random damage is evaluated using a growth model for individual damages which can be obtained from experimental data. The distribution function of damage intensity after inspection and repair is updated using a Bayesian analysis that takes the quality of inspection into account. With this posterior distribution, the time-dependent strength after the inspection and repair can be evaluated. Time-dependent reliability analysis can be used to determine inspection/repair strategies that are necessary to keep the failure probability of the component lower than an established target. (author)

  5. 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

  6. After-production and in-service inspections of components of nuclear power plant primary coolant circuits

    International Nuclear Information System (INIS)

    Slama, K.; Svetlik, M.

    1990-01-01

    A new diagnostic system was developed for detecting defects in the material of mechanically loaded equipment. It is based on the measurement of elastic strain waves propagating through the materials. The instrument units as well as the methodology and software are of Czechoslovak origin and can be modified to conform to the requirements and experience of the user. The way of applying the method to the diagnostics of pressure vessels, main circulation pumps of the pressurizers and of the primary piping is described. Some results of after-production and in-service acoustic emission tests are given, as are the technical parameters of the acoustic emission analyzer. (M.D.). 5 figs

  7. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    Holman, G.S.

    1989-12-01

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  8. Overview of improvements in work practices and instrumentation for CANDU primary heat transport feeders in-service inspections

    Energy Technology Data Exchange (ETDEWEB)

    Marcotte, O., E-mail: olivier@nucleom.ca [Nucleom Inc., Quebec, Quebec (Canada); Rousseau, G., E-mail: rousseau.gilles.a@hydro.qc.ca [Hydro Quebec, Becancour, Quebec (Canada); Rochefort, E., E-mail: erochfort@zetec.com [Zetec Canada, Quebec, Quebec (Canada)

    2013-01-15

    The Canadian nuclear industry has developed many advanced non-destructive inspection techniques to be applied safely in hazardous environments in recent years. Automated systems, manual tooling and specialized software modules have been designed since early 2000s to provide complete and very efficient fitness for service inspection of primary heat transport system carbon steel feeder pipes. These techniques deal with complex geometries, difficult access and, radioactive environment. Complementary NDE techniques, namely Ultrasounds, eddy current, phased-array UT and automated scanners are used. This presentation describes the improvements in inspection practices and the advanced data analysis features. (author)

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

    International Nuclear Information System (INIS)

    Jordan, G.M.

    1977-01-01

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

  10. 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.

  11. 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.)

  12. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

    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

  14. 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)

  15. Automated ultrasonic testing of nuclear reactor welds and overlays in pre-service and in-service inspections

    International Nuclear Information System (INIS)

    Sladky, J.

    1988-01-01

    Since 1982, automatic pre-service and in-service checks are being made of welded joints and overlays on pressure vessels of WWER-440 nuclear reactors in Czechoslovakia. This is being done using the SKODA REACTORTEST TRC facility which is used for checking peripheral welded joints on the pressure vessel, neck joints, overlays in other selected areas of the cylindrical section of the pressure vessel, on radius transitions of the pressure vessel and of necks, and on the cylindrical part of necks, and also for checking the base material in selected parts of the pressure vessel and the base material of the neck extension piece. The tests are of two types, namely tests of peripheral welds and overlays of the cylindrical parts of the pressure vessel, and tests of the necks. Different ultrasonic probe holders are used for the tests, with totally different design. Ultrasonic probes which were initially used were of foreign make while at present, those of Czechoslovak make are used. For each pressure vessel a set of ultrasonic probes is used which should suffice for the life of the vessel. Experience gained so far is being used in work on the project of a new device for testing nuclear reactor presure vessels from the inside. (Z.M.)

  16. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  17. 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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  19. Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

    International Nuclear Information System (INIS)

    Comby, R.

    1997-01-01

    Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection. (orig.)

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

    International Nuclear Information System (INIS)

    Doctor, S.R.; Reid, L.D.; Hall, T.E.; Littlefield, R.J.; Gilbert, R.W.; Crawford, S.L.; Baldwin, A.J.; Bowey, R.E.

    1985-10-01

    A multi-year program is underway at Pacific Northwest Laboratory (PNL) to move the synthetic aperture focusing technique from the Laboratory into the field to inspect light water reactor components. This report is a summary of highlights from the third year's efforts. The work presented here includes: scanner development, SAFT-UT signal processing techniques, SAFT-UT graphics package development, SAFT-UT real-time processor, SAFT-UT field system integration, SAFT-UT evaluation on CCSS, a field trip demonstrating in-field SAFT data processing, and future work. 11 figs

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. 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)

  3. Analysis of in-service inspection guides for post-tensioning systems in containment structures with greased tendons

    International Nuclear Information System (INIS)

    Bae, I. H.; Choi, I. G.; Seo, J. M.

    2000-01-01

    Prestressed concrete containments(PCC) are inspected periodically to ensure structural integrity and to identify and correct problems before they come critical. These inspections are conducted in accordance with the Nuclear Regulatory Committee(NRC) Regulatory Guide. As experience with the surveillance mounts, the guide is revised to keep pace with technological advances in containment design and to reflect the knowledge obtained experience. In the study, an analysis of the available utility surveillance data and an evaluation of the NRC Regulatory Guide and the ASME Code are conducted. The results indicate that the average value of the lower bound defined in Regulatory Guide 1.35.1 can be less than the minimum required force level at 40 years for the existing containment and thus an acceptable alternative has to be presented to meet the allowance of the guide. Tendons in excess of the upper bound or 70% of the ultimate tensile strength of the tendon are observed. Such an occurrence indicates tendon behavior outside the tolerance band that expected, and the cause of such behavior has to be investigated. It is shown that the behavior of the predicted prestressing force with time based on assumed losses does quite different from that of the measured tendon force with time and, therefore, studies for the significant factors, such as shrinkage and creep of concrete, that influence the time-dependent losses in PCC are needed. Additional research on the long-term effects of the impurity levels on the effectiveness of the grease seems justified

  4. 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.)

  5. 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)

  6. 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)

  7. Progress toward risk informed regulation

    International Nuclear Information System (INIS)

    Rogers, K.C.

    1997-01-01

    For the last several years, the NRC, with encouragement from the industry, has been moving in the direction of risk informed regulation. This is consistent with the regulatory principle of efficiency, formally adopted by the Nuclear Regulatory Commission in 1991, which requires that regulatory activities be consistent with the degree of risk reduction they achieve. Probabilistic risk analysis has become the tool of choice for selecting the best of several alternatives. Closely related to risk informed regulation is the development of performance based rules. Such rules focus on the end result to be achieved. They do not specify the process, but instead establish the goals to be reached and how the achievement of those goals is to be judged. The inspection and enforcement activity is based on whether or not the goals have been met. The author goes on to offer comments on the history of the development of this process and its probable development in the future. He also addresses some issues which must be resolved or at least acknowledged. The success of risk informed regulation ultimately depends on having sufficiently reliable data to allow quantification of regulatory alternatives in terms of relative risk. Perhaps the area of human reliability and organizational performance has the greatest potential for improvement in reactor safety. The ability to model human performance is significantly less developed that the ability to model mechanical or electrical systems. The move toward risk informed, performance based regulation provides an unusual, perhaps unique, opportunity to establish a more rational, more effective basis for regulation

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. 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)

  10. Methodology for qualification of in-service inspection systems for WWER nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1998-03-01

    Program was initiated by IAEA in 1990 with the aim to assist the countries of Central and Eastern Europe and former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives were: to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the competence and and adequacy of safety improvement programs. The scope was extended in 1992 to include RBMK, WWER-440/312 and WWER-1000 plants in operation and under construction. Integrity of primary circuit is fundamental for the safe operation of any nuclear power plant. In-service inspection (ISI) in general terms and in particular, non-destructive tests (NDT) play a key role in maintaining primary circuit integrity. This report provides a methodology for qualification of ISI systems which might be used by WWER operating countries as a commonly accepted basis for further development of the necessary qualification related infrastructures. It also provides several qualification principles defining the administrative framework needed for the practical implementation of the methodology, a description of the process of qualification of an inspection system, specifying its minimum technical and documentation related requirements, as well as specific requirements with regard to the NDT procedures, equipment and personnel to be qualified and the test specimen to be used in practical trials. Finally, the report suggests an appropriate distribution of responsibilities among all the parties involved in a qualification process, based on international practice

  11. 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)

  12. 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

  13. Capability evaluation of Eddy current and ultrasonic in-service inspections of steam generator tubes. A status report of PISC III Action 5

    International Nuclear Information System (INIS)

    Bieth, M.; Birac, C.; Comby, R.

    1998-01-01

    Document summarizes the PISC III (Programme for the Inspection of Steel Components) report No. 41, full description of the PISC III Action 5 on Steam Generator Tubes Inspection, containing all details and final conclusions which are still to be approved by the PISC III Management Board. The report was prepared by the reference laboratory of PISC under guidance and with continuous contribution of the members of the Data Analysis Group (DAG) of this PISC III. There were several procedures which demonstrated good detection capability of major flaws in typical locations of the steam generator. Conclusions of the exercise indicate that capability demonstration is necessary to qualify in service inspection procedures for steam generator tubes

  14. Study on in-service inspection and repair program and related plant design for Japan Sodium-Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Suzuki, Shinichi; Kotake, Shoji; Nishiyama, Noboru; Uzawa, Masayuki

    2011-01-01

    Maintenance and repair program and conformity with them were investigated as a part of the conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR). The maintenance program was set by taking the feature of sodium-cooled reactors and domestic practice of LWRs into account. Both of regulatory required inspection and voluntary inspection, which are conducted in the domestic LWRs, were counted. The regulatory required ISI program was based on that of the previous Japanese SFRs, LWRs (JSME S NA1) and liquid metal cooled reactors (ASME section XI division 3). Parts to be inspected, methods of inspection were identified for major structures and components. Concerning the repair program, we set three levels of repair requirements based on estimated frequency of defect and failure during the plant life time. For level 1, which might be occur several times during the plant life time, it is required to be easily repaired in a short period. Access routes and working space are considered in the component design and its arrangement. For level 2, which might be unlikely to occur during the plant life time, it is required to check that the repair work is feasible in a practical time range. For level 3, which frequency is negligible small, repair is not taken into account but the feasibility was investigated. The plant design shall be done so that all of above mentioned inspection and repair can be conducted. It is desired to ensure accessibility for all of the coolant and cover gas boundaries and the internal structures in order to cope with unforeseen troubles. Access routes for the reactor vessel and its internal structures, piping, pumps and intermediate heat exchangers and steam generators were investigated. As the results of that, possible ways for implementation of the maintenance and repair were identified. (author)

  15. 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)

  16. 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)

  17. Preventive and Predictive Maintenance, Warehousing of Spares, Periodic Testing and In-Service Inspection Activities at the Nigerian Research Reactor-1 Facility

    Energy Technology Data Exchange (ETDEWEB)

    Yusuf, I.; Mati, A. A.; Dewu, B. B.M., [Centre for Energy Research and Training, Ahmadu Bello University, Zaria (Nigeria)

    2014-08-15

    The Nigerian Research Reactor–1, or NIRR-1, is sited at Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. Activities on preventive or routine maintenance have been institutionalized since the commissioning of the reactor in February 2004. This has grossly reduced the rates of corrective maintenance activities and helped the reactor management a great deal in predicting failure rates of reactor components and other auxiliary units. Routine maintenance of systems and components are being carried out on a weekly, quarterly and annual basis based on manufacturer’s recommendations, which have been reviewed and improved over the years. The paper presents the implementation of maintenance activities in NIRR-1 from its initial criticality in 2004 till today and the new scheme for periodic testing and in-service-inspection developed after an IAEA Integrated Safety Assessment of Research Reactors mission. The measures put in place are envisaged to reduce the negative impact of ageing on NIRR-1 and its auxiliary systems. (author)

  18. 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

  19. Risk Informed Structural Systems Integrity Management

    DEFF Research Database (Denmark)

    Nielsen, Michael Havbro Faber

    2017-01-01

    The present paper is predominantly a conceptual contribution with an appraisal of major developments in risk informed structural integrity management for offshore installations together with a discussion of their merits and the challenges which still lie ahead. Starting point is taken in a selected...... overview of research and development contributions which have formed the basis for Risk Based Inspection Planning (RBI) as we know it today. Thereafter an outline of the methodical basis for risk informed structural systems integrity management, i.e. the Bayesian decision analysis is provided in summary....... The main focus is here directed on RBI for offshore facilities subject to fatigue damages. New ideas and methodical frameworks in the area of robustness and resilience modeling of structural systems are then introduced, and it is outlined how these may adequately be utilized to enhance Structural Integrity...

  20. Development and trial application of risk information

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been doing various activities to stimulate the introduction of Risk Informed Regulation (RIR) to the safety regulation of nuclear power plants (NPPs) in Japan. Some applications are already incorporated, such as the regulatory review of Maintenance Programs and Safety Significance Evaluation for Inspection Findings. In consideration with the experience of the accident in Fukushima Daiichi Nuclear Power Station, JNES addressed development of regulatory guidelines, evaluation of the current condition of Fukushima Daiichi Nuclear Power Station, evaluation of effectiveness of severe accident management measures with the backgrounds of insights and experiences on probabilistic risk assessment (PRA) and RIR. Especially, the experiences were applied to the development of the methodologies for evaluation of effectiveness of severe accident managements. As for inspection and operation of NPPs, JNES enhanced the PRA scope applied to the importance analysis for Maintenance Program, SDP and RI-ISI in consideration with the insights of RIR in Japan and other countries. (author)

  1. 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

  2. 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)

  3. In-service inspection of electrical items

    International Nuclear Information System (INIS)

    Sastry, M.S.N.

    2002-01-01

    Power Supply and Distribution System caters to electrical power requirements of all drives and other equipment. The system also provides control power. Main equipment in the system are Transformers, Switch gear, cables, motors, emergency D- G sets, UPS, and storage battery banks. The talk is covered under the following broad heads: 1. Brief description of typical power supply system, 2. Insulation checks in LV, MV, and HV systems, 3. Monitoring the insulating oil of transformer, 4. Eddy current checking of the motor cores, 5. Checking of cable and bus bar joints, 6. Periodic check on batteries

  4. 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)

  5. Integrated Risk Information System (IRIS)

    Data.gov (United States)

    U.S. Environmental Protection Agency — EPA?s Integrated Risk Information System (IRIS) is a compilation of electronic reports on specific substances found in the environment and their potential to cause...

  6. Proceedings of the IAEA Specialists' Meeting on experience and further improvement of in-service inspection methods and programmes of NPPs with particular emphasis on on-line techniques

    International Nuclear Information System (INIS)

    1989-01-01

    The proceedings are presented of an IAEA specialists meeting devoted to the exchange of views and experience with respect to understanding and quantifying materials ageing and degradation processes, to assessing residual life of components, to in-service materials testing, all this with the objective of improving reliability and, wherever possible, extending the useful life of nuclear power plants. (L.O.)

  7. 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

  8. 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

  9. Risk informed decision making - a pre-study

    International Nuclear Information System (INIS)

    Simola, K.; Pulkkinen, U.

    2004-04-01

    Examples of risk-informed decisions are establishing maintenance programmes, optimising inspection policies and justifying plant modifications, and revising technical specifications. Applications in daily situations can be such as accepting or rejecting exemptions from technical specifications. The aim of this pre-study was to identify the status of risk-informed decision making at Swedish and Finnish nuclear power plants and nuclear safety authorities. Responses to a questionnaire were obtained either by interviews or by e-mail from two Swedish and two Finnish NPPs, SKI and STUK. The development of a risk-informed decision procedure based on decision analytic ideas is worth recommending. A clear documentation format is a part of such procedure. In order to serve as a basis for final decision, the documentation should include clearly defined decision criteria, qualification of PSA model for the issue under analysis, description of most important uncertainties and assumptions. (au)

  10. 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)

  11. 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)

  12. 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)

  13. 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)

  14. 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.)

  15. Communicating risk information and warnings

    Science.gov (United States)

    Mileti, D. S.

    1990-01-01

    Major advances have occurred over the last 20 years about how to effectively communicate risk information and warnings to the public. These lessons have been hard won. Knowledge has mounted on the finding from social scientific studies of risk communication failures, successes and those which fell somewhere in between. Moreover, the last 2 decades have borne witness to the brith, cultivation, and blossoming of information sharing between those physical scientists who discover new information about risk and those communcation scientists who trace its diffusion and then measure pbulic reaction. 

  16. Air Risk Information Support Center

    Energy Technology Data Exchange (ETDEWEB)

    Shoaf, C.R.; Guth, D.J. [Environmental Protection Agency, Washington, DC (United States)

    1990-12-31

    The Air Risk Information Support Center (Air RISC) was initiated in early 1988 by the US Environmental Protection Agency`s (EPA) Office of Health and Environmental Assessment (OHEA) and the Office of Air Quality Planning and Standards (OAQPS) as a technology transfer effort that would focus on providing information to state and local environmental agencies and to EPA Regional Offices in the areas of health, risk, and exposure assessment for toxic air pollutants. Technical information is fostered and disseminated by Air RISCs three primary activities: (1) a {open_quotes}hotline{close_quotes}, (2) quick turn-around technical assistance projects, and (3) general technical guidance projects. 1 ref., 2 figs.

  17. 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.

  18. 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.

  19. Launcher In-Service Workspace

    Data.gov (United States)

    Federal Laboratory Consortium — Purpose:Through Engineering Investigations (EIs), testing, development, evaluation and system integration of Aircraft Launch and Recovery Equipment (ALRE) in-service...

  20. Integrated risk information system (IRIS)

    Energy Technology Data Exchange (ETDEWEB)

    Tuxen, L. [Environmental Protection Agency, Washington, DC (United States)

    1990-12-31

    The Integrated Risk Information System (IRIS) is an electronic information system developed by the US Environmental Protection Agency (EPA) containing information related to health risk assessment. IRIS is the Agency`s primary vehicle for communication of chronic health hazard information that represents Agency consensus following comprehensive review by intra-Agency work groups. The original purpose for developing IRIS was to provide guidance to EPA personnel in making risk management decisions. This original purpose for developing IRIS was to guidance to EPA personnel in making risk management decisions. This role has expanded and evolved with wider access and use of the system. IRIS contains chemical-specific information in summary format for approximately 500 chemicals. IRIS is available to the general public on the National Library of Medicine`s Toxicology Data Network (TOXNET) and on diskettes through the National Technical Information Service (NTIS).

  1. 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.

  2. 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

  3. Implementing risk-informed life-cycle design

    International Nuclear Information System (INIS)

    Hill, Ralph S. III

    2007-01-01

    This paper describes a design process based on risk-informed probabilistic methodologies that cover a facility's life-cycle from start of conceptual design through decontamination and decommissioning. The concept uses probabilistic risk assessments to identify target reliabilities for facility systems and components. Target reliabilities are used in system and subsystem simulation analyses to determine the optimum combination of initial system and component construction reliability, maintenance frequency, and inspection frequency for both active and passive components. The target reliabilities are also used for system based code margin exchange to reduce excessive level of margins to appropriate levels resulting in a more flexible structure of codes and standards that improves facility reliability and cost. The paper includes a description of a risk informed life-cycle design process, a summary of work being done, and a discussion of work needed to implement the process. (author)

  4. 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.

  5. 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

  6. 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)

  7. Issues and future directions of 'introduction of Risk Informed Regulation'

    International Nuclear Information System (INIS)

    2008-01-01

    In 2004, Nuclear Safety Commission of Japan (NSC) set up a taskforce for developing a scheme of Risk Informed Regulation (RIR) introduction, which had discussed various aspects of utilization of risk information in nuclear regulations. In the final report compiled in 2007, the taskforce identified the significance of RIR introduction from viewpoint of enhancement of rationality, consistency and transparency as well as appropriate allocation of regulatory resources in the nuclear safety regulation and major issues for the further promotion were as follows: NSC is to maintain the comprehensively promoting function of the RIR introduction with reviewing global trends and latest technological knowledge, regulatory bodies are to further enhance the RIR in inspection and operational area, utilities and industries are to utilize further risk information in their self-maintenance activities and to gather trouble information data, academic societies are to establish the standards for risk evaluation technology and RIR, and research institutes are to carry out safety research and risk training for risk-communicators. This article presented summary of the final report aiming at upgrading safety regulation with enhancement of RIR introduction. (T. Tanaka)

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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

  15. Report on probabilistic safety assessment (PSA) quality assurance in utilization of risk information

    International Nuclear Information System (INIS)

    2006-12-01

    Recently in Japan, introduction of nuclear safety regulations using risk information such as probabilistic safety assessment (PSA) has been considered and utilization of risk information in the rational and practical measures on safety assurance has made a progress to start with the operation or inspection area. The report compiled results of investigation and studies of PSA quality assurance in risk-informed activities in the USA. Relevant regulatory guide and standard review plan as well as issues and recommendations were reviewed for technical adequacy and advancement of probabilistic risk assessment technology in risk-informed decision making. Useful and important information to be referred as issues in PSA quality assurance was identified. (T. Tanaka)

  16. 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

  17. Use of risk information in regulatory reviews

    International Nuclear Information System (INIS)

    Sagar, B.; Benke, R.; Mohanty, S.

    2004-01-01

    The regulatory framework for licensing any high-level waste repository at Yucca Mountain in the United States, calls for appropriate use of risk information to ensure operational safety during the pre-closure period and long-term safety during the post-closure period. This paper focuses on the post-closure period. Regulations in the Code of Federal Regulations (CFR), Title 10, Part 63, apply to any repository at Yucca Mountain and envision use of probabilistic methods to develop quantitative risk information. Accumulated engineering and scientific experience at Yucca Mountain and analog sites and quantitative risk information from studies conducted by the implementer, regulator, and others are combined to formulate 'risk insights,' which are then used to plan and execute regulatory reviews. The U.S. Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA) recently consolidated the knowledge gained during several g ears and developed such risk insights for the potential repository at Yucca Mountain. This paper discusses the types of risk information used to generate risk insights and how the risk insights will be used in regulatory reviews. A companion paper presents more details on sensitivity analysis methods used to generate risk information. (authors)

  18. Regulatory environment of transitioning to risk-informed regulations in U.S.A

    International Nuclear Information System (INIS)

    Choi, C. H.; Kim, C. H.

    1999-01-01

    With the publication of the PRA Policy Statement and recent regulatory guides, the U.S.NRC makes a continuous approach towards risk-informed regulations with the goal of establishing an overall framework for risk-informed decisions in all regulatory activities as well as plant specific licensing issues. Faced with the changing environment of deregulation of the electricity generation market, the licensee's effort to reduce design margins to enhance flexibility and to relieve unnecessary regulatory burdens have been focused on the control and reduction of plant operating costs. The risk-informed approach provides a structured, systematic, and defensible method that can be applied not only to rulemaking, but also to licensing, inspection, enforcement, and performance assessment, as well as provides basis for prioritization in the establishment of programs and the allocation of resources. This report describes the current regulatory environment of transitioning to risk-informed regulations with an emphasis on its background, concepts, regulatory guides, proposed options for modifying the 10CFR50, and risk-informed applications in U.S.A. Review of the risk informed applications utilizing the information provided by the PRAs and their insights in the U.S.NRC and nuclear industry will provide the insights of predicting the expected regulation changes in Korea. Also it could provide the applicable methods or guides for the implementation of the risk-informed applications in plant design and operations. (author)

  19. 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)

  20. 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)

  1. 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

  2. Risk informed decisions and regulations - STUK's policy and current practice

    International Nuclear Information System (INIS)

    Julin, A.; Niemalae, I.; Virolainen, R.

    2001-01-01

    Consideration of severe accidents beyond the traditional design basis, including full core melt accidents, has become an important ingredient of regulatory process in Finland. Accordingly, plant-specific level-1 and level-2 PSA studies are a regulatory requirement. These studies are being used in a living fashion both at the utilities and STUK. Plant specific living PSAs have been completed for all operating Finnish plants, including internal initiators, fires, flooding, harsh weather conditions seismic events for operation mode and internal events for low power mode. Many specific applications of the Living PSA have already been introduced but some are still waiting for further development such as Risk Informed ISI, IST and Tech Specs. Examples of safety issues, for which the PSA insights give an improved basis for decisions, are approvals of plant modifications and resolution of testing, inspection and maintenance strategies. PSA insights are also of value in assessing meaningfulness of requirements which are based on traditional engineering judgement but do not form an essential part of defence-in-depth concept. Examples of such requirements are details of safety classification and many Technical Specification requirements. STUK has recently conducted a pilot study on risk-informed ISI. The aim of the study was to explore how the plant specific PSAs could best be used for assessment of the ISI programmes. This paper discusses the findings obtained during the pilot study on risk-informed ISI of pipings. The study produced essential insights of the applied method. Furthermore, the study gave guidance to extract items for further development. Based on these results and overall experience the general suitability of the method for further application is evaluated. (author)

  3. Implementing risk-informed life-cycle design

    International Nuclear Information System (INIS)

    Hill, Ralph S.

    2009-01-01

    This paper describes a design process based on risk-informed probabilistic design methodologies that cover a facility's life-cycle from start of conceptual design through decontamination and decommissioning. The concept embodies use of probabilistic risk assessments to establish target reliabilities for facility systems and components. The target reliabilities are used for system based code margin exchange and performance simulation analyses to optimize design over all phases (design, construction, operation and decommissioning) of a facility's life-cycle. System based code margin exchange reduces excessive level of construction margins for passive components to appropriate levels resulting in a more flexible structure of codes and standards that improves facility reliability and cost. System and subsystem simulation analyses determine the optimum combination of initial system and component construction reliability, maintenance frequency, and inspection frequency for both active and passive components. The paper includes a description of these risk-informed life-cycle design processes, a summary of work being done, and a discussion of additional work needed to implement the process.

  4. Crack Characterisation for In-service Inspection Planning - An Update

    Energy Technology Data Exchange (ETDEWEB)

    Waale, Jan [lnspecta Technology AB, Stockholm (Sweden)

    2006-05-15

    One important factor to optimize the NDT equipment and NDT procedure is to know the characteristics of the specific defects being sought for in each case. Thus, access is necessary to reliable morphology data of defects from all possible degradation mechanisms in all existing materials of the components that are subject to the NDT. In 1994 the Swedish Nuclear Power Inspectorate (SKI) initiated a project for compiling crack morphology data based on systematic studies of cracks that have been observed in different plants (nuclear and non-nuclear) in order to determine typical as well as more extreme values of e.g. orientation, width and surface roughness. Although, a large number of identified cracking incidents was covered by the work it was recognised that further studies were needed to increase the data base, and thereby getting more confidence in the use of different crack characteristic data for NDT development and qualification purposes. That is the major reason why the present work was initiated. A thorough review of the SKI archives was performed aiming to find useful material from the time period between 1994 and today to compile complementary data and produce an update. Furthermore, older material was collected and evaluated. Thus, the data cover cracking found within the time period 1977-2003. In addition, useful material was supplied by the Swedish nuclear power plants. The evaluation and presentation of the results are similar to the 1994 study, with a few exceptions. The base for the evaluation is failure analysis reports, where the crack morphology parameters were measured from photos on cracked surfaces or cross sections through cracks. The resulting data were divided into seven groups depending on the cracking mechanism/material group combination. The data groups are: IGSCC in austenitic stainless steels; IGSCC in nickel base alloys; IDSCC in nickel base weld metal; TGSCC in austenitic stainless steels Thermal fatigue in austenitic stainless steels; Mechanical fatigue; and Solidification cracking in weld metal. The evaluated parameters were divided into visually detectable and metallurgical parameters, which need to be evaluated from a cross-section. The visually detectable parameters are; location, orientation and shape in surface direction and finally the number of cracks in the cracked region. The metallurgical parameters are; orientation and shape in the through thickness direction, macroscopic branching, crack tip radius, crack surface roughness, crack width and finally discontinuous appearance. The morphology parameters were statistically processed and the results are presented as minimum, maximum. mean, median and scatter values for each data group, both in tables and in various graphs. Finally each morphology parameter is compared between the seven data groups. A brief description of typical characteristics of each data group is given below. Most IGSCC develop next to welds with straight or winding cracks oriented almost parallel to the weld. Single cracking is most common but occasionally two cracks are formed on each side of the weld. In the through thickness direction IGSCC is typically winding or lightly bend and macroscopic branching is rare. The surface roughness is normally on a grain size magnitude and the cracks are particularly narrow providing secondary corrosion is small. Similar characteristics to IGSCC in austenitic stainless steels may be expected. However, cracking close to weld are less frequent and macroscopic branching is more common for IGSCC in nickel base alloys compared to austenitic stainless steels. Typically IDSCC is winding or straight, single cracking in the weld metal transverse to the weld. In the through thickness direction IDSCC cause typically winding, non branched cracks with large surface roughness due to course solidification microstructure. The crack width often shows large variation along the crack and a width close to zero at the surface intersection is common. Typically, TGSCC is branched both in surface and through thickness direction. The crack orientation shows a random distribution and the number of cracks is large. The crack surface roughness show low values and the crack width is typically medium range compared with the other groups. A large number of randomly oriented cracks are typical for thermal fatigue. However, single or few cracks with similar orientation occur. In the through thickness direction straight, non-branched cracking oriented in right angle to the surface is most common. The crack surface roughness is of medium range and larger than for mechanical fatigue. Typically straight, single cracking oriented parallel with stress raisers is common for mechanical fatigue. In the through thickness direction most cracks are straight, non branched and oriented in right angle to the surface. The crack surface roughness is the smallest and the correlation length the highest of all groups. Solidification cracks occur equally frequent parallel as well as transversal to the weld. A large number of cracks are common. In the through thickness direction the cracks seldom show branching and is most often oriented close to 90 deg to the surface. The crack surface roughness is in the medium range and far below the one for IDSCC, which was not expected.

  5. Crack Characterisation for In-service Inspection Planning - An Update

    International Nuclear Information System (INIS)

    Waale, Jan

    2006-05-01

    One important factor to optimize the NDT equipment and NDT procedure is to know the characteristics of the specific defects being sought for in each case. Thus, access is necessary to reliable morphology data of defects from all possible degradation mechanisms in all existing materials of the components that are subject to the NDT. In 1994 the Swedish Nuclear Power Inspectorate (SKI) initiated a project for compiling crack morphology data based on systematic studies of cracks that have been observed in different plants (nuclear and non-nuclear) in order to determine typical as well as more extreme values of e.g. orientation, width and surface roughness. Although, a large number of identified cracking incidents was covered by the work it was recognised that further studies were needed to increase the data base, and thereby getting more confidence in the use of different crack characteristic data for NDT development and qualification purposes. That is the major reason why the present work was initiated. A thorough review of the SKI archives was performed aiming to find useful material from the time period between 1994 and today to compile complementary data and produce an update. Furthermore, older material was collected and evaluated. Thus, the data cover cracking found within the time period 1977-2003. In addition, useful material was supplied by the Swedish nuclear power plants. The evaluation and presentation of the results are similar to the 1994 study, with a few exceptions. The base for the evaluation is failure analysis reports, where the crack morphology parameters were measured from photos on cracked surfaces or cross sections through cracks. The resulting data were divided into seven groups depending on the cracking mechanism/material group combination. The data groups are: IGSCC in austenitic stainless steels; IGSCC in nickel base alloys; IDSCC in nickel base weld metal; TGSCC in austenitic stainless steels Thermal fatigue in austenitic stainless steels; Mechanical fatigue; and Solidification cracking in weld metal. The evaluated parameters were divided into visually detectable and metallurgical parameters, which need to be evaluated from a cross-section. The visually detectable parameters are; location, orientation and shape in surface direction and finally the number of cracks in the cracked region. The metallurgical parameters are; orientation and shape in the through thickness direction, macroscopic branching, crack tip radius, crack surface roughness, crack width and finally discontinuous appearance. The morphology parameters were statistically processed and the results are presented as minimum, maximum. mean, median and scatter values for each data group, both in tables and in various graphs. Finally each morphology parameter is compared between the seven data groups. A brief description of typical characteristics of each data group is given below. Most IGSCC develop next to welds with straight or winding cracks oriented almost parallel to the weld. Single cracking is most common but occasionally two cracks are formed on each side of the weld. In the through thickness direction IGSCC is typically winding or lightly bend and macroscopic branching is rare. The surface roughness is normally on a grain size magnitude and the cracks are particularly narrow providing secondary corrosion is small. Similar characteristics to IGSCC in austenitic stainless steels may be expected. However, cracking close to weld are less frequent and macroscopic branching is more common for IGSCC in nickel base alloys compared to austenitic stainless steels. Typically IDSCC is winding or straight, single cracking in the weld metal transverse to the weld. In the through thickness direction IDSCC cause typically winding, non branched cracks with large surface roughness due to course solidification microstructure. The crack width often shows large variation along the crack and a width close to zero at the surface intersection is common. Typically, TGSCC is branched both in surface and through thickness direction. The crack orientation shows a random distribution and the number of cracks is large. The crack surface roughness show low values and the crack width is typically medium range compared with the other groups. A large number of randomly oriented cracks are typical for thermal fatigue. However, single or few cracks with similar orientation occur. In the through thickness direction straight, non-branched cracking oriented in right angle to the surface is most common. The crack surface roughness is of medium range and larger than for mechanical fatigue. Typically straight, single cracking oriented parallel with stress raisers is common for mechanical fatigue. In the through thickness direction most cracks are straight, non branched and oriented in right angle to the surface. The crack surface roughness is the smallest and the correlation length the highest of all groups. Solidification cracks occur equally frequent parallel as well as transversal to the weld. A large number of cracks are common. In the through thickness direction the cracks seldom show branching and is most often oriented close to 90 deg to the surface. The crack surface roughness is in the medium range and far below the one for IDSCC, which was not expected

  6. In-service inspection of pool type research reactors

    International Nuclear Information System (INIS)

    Rajamani, K.

    2002-01-01

    In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

  7. In-service ultrasonic inspection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Prepechal, J.; Sulc, J.

    1982-01-01

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

  8. 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

  9. Inservice testing of vertical pumps

    International Nuclear Information System (INIS)

    Cornman, R.E. Jr.; Schumann, K.E.

    1994-01-01

    This paper focuses on the problems that may occur with vertical pumps while inservice tests are conducted in accordance with existing American Society of Mechanical Engineers Code, Section XI, standards. The vertical pump types discussed include single stage, multistage, free surface, and canned mixed flow pumps. Primary emphasis is placed on the hydraulic performance of the pump and the internal and external factors to the pump that impact hydraulic performance. In addition, the paper considers the mechanical design features that can affect the mechanical performance of vertical pumps. The conclusion shows how two recommended changes in the Code standards may increase the quality of the pump's operational readiness assessment during its service life

  10. Psychodrama Techniques for Inservice Teacher Training

    Science.gov (United States)

    Kohut, Sylvester, Jr.

    1976-01-01

    By using psychodrama techniques in inservice programming, improvement in communication among members of the school community occurs. With modification the psychodrama approach can be used with inservice teachers and administrators to improve self-realization and communication. A program using psychodrama techniques is described in the article.…

  11. 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)

  12. 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)

  13. 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.

  14. Risk - Informed decision making at Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    Vaurio, J.K. [Fortum Power and Heat Oy, Loviisa (Finland)

    1999-09-01

    PSA has been used in many ways for risk-informed decision making at Loviisa power station. The most fruitful areas so far include: 1) Identification of dominating risk contributors and possible means for reducing risk by plant modification and improved procedures. 2) Providing risk perspective and economic criteria for assessing backfitting proposals. 3) Assessing the significance of ageing and needs for renewals. 4) Limiting, prioritising and optimising plant modifications. 5) Reducing testing requirements. 6) Justification of temporary aswell as permanent configurations and extended outage times. 7) Planning and prioritisation of training programs. (au)

  15. Risk - Informed decision making at Loviisa NPP

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1999-01-01

    PSA has been used in many ways for risk-informed decision making at Loviisa power station. The most fruitful areas so far include: 1) Identification of dominating risk contributors and possible means for reducing risk by plant modification and improved procedures. 2) Providing risk perspective and economic criteria for assessing backfitting proposals. 3) Assessing the significance of ageing and needs for renewals. 4) Limiting, prioritising and optimising plant modifications. 5) Reducing testing requirements. 6) Justification of temporary as well as permanent configurations and extended outage times. 7) Planning and prioritisation of training programs. (au)

  16. 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)

  17. Moving along the risk-informed path - drivers and challenges

    International Nuclear Information System (INIS)

    Petrella, S.; Dermarkar, F.; Austman, D.; Chun, R.

    2004-01-01

    The use of risk criteria and information has always been an important component of the licensing of nuclear reactors in Ontario. Early applications included the development of Safety Design Matrices in the early 80's for more comprehensive design reviews, the use of the Darlington Probabilistic Safety Evaluation in the late-80's to support the C-6 event categorization process for the initial Darlington licensing, and the use of risk-insights, along with cost benefit criteria, to evaluate design options for the Pickering A special safety system design retrofits and modifications, such as Shutdown System (SDS) enhancement and Emergency Coolant Injection System (ECIS) upgrades. Operational uses have included the management of outage risk, the review of containment test frequencies, and the review of abnormal operating configurations such as maintenance of electrical power supplies. It is envisaged that such applications will continue, aided by the continued development of Probabilistic Risk Assessments techniques. To provide enhanced assurance of safety, to evaluate current and new safety-related issues as they arise, and to realize the full safety and economic benefits of risk-insights, OPG and Bruce Power have, along with their Industry partners, embarked on a path to further increase the use of risk-information in the operation and licensing of their reactors. Drivers include the resolution of long standing safety analysis issues surrounding LBLOCA analyses and generic action items, the application of risk-insights to guide the response to design, inspection, and analysis findings, and the optimization of operating, maintenance, and testing programs. As well, a major near term driver is input to plant refurbishment decisions. In moving to a more risk-informed approach to safety assessment, challenges include developing and implementing the required suite of concepts, tools, processes and criteria, and reconciling the current largely deterministic approach to safety

  18. Study on the risk-informed regulation of NPP

    International Nuclear Information System (INIS)

    Wang Chaogui

    2007-01-01

    The risk-informed regulation is a modern type of NPP safety management mode using both deterministic and probabilistic approaches. It is necessary to entirely and systematically study the associated regulations, standards and practices in order to promote the developments of risk-informed regulations in China. This paper introduces the risk-informed regulation, gives out the basic principles, method and acceptance risk criteria of risk-informed decision,making, discusses the PSA requirements for risk-informed decision-making and makes some suggestions about the application of risk-informed regulations in Chinese NPP. (authors)

  19. 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)

  20. 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.)

  1. NASA Risk-Informed Decision Making Handbook

    Science.gov (United States)

    Dezfuli, Homayoon; Stamatelatos, Michael; Maggio, Gaspare; Everett, Christopher; Youngblood, Robert; Rutledge, Peter; Benjamin, Allan; Williams, Rodney; Smith, Curtis; Guarro, Sergio

    2010-01-01

    This handbook provides guidance for conducting risk-informed decision making in the context of NASA risk management (RM), with a focus on the types of direction-setting key decisions that are characteristic of the NASA program and project life cycles, and which produce derived requirements in accordance with existing systems engineering practices that flow down through the NASA organizational hierarchy. The guidance in this handbook is not meant to be prescriptive. Instead, it is meant to be general enough, and contain a sufficient diversity of examples, to enable the reader to adapt the methods as needed to the particular decision problems that he or she faces. The handbook highlights major issues to consider when making decisions in the presence of potentially significant uncertainty, so that the user is better able to recognize and avoid pitfalls that might otherwise be experienced.

  2. 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)

  3. 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.

  4. 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

  5. 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

  6. 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.

  7. Inspection Database

    Data.gov (United States)

    U.S. Department of Health & Human Services — FDA is disclosing the final inspection classification for inspections related to currently marketed FDA-regulated products. The disclosure of this information is not...

  8. 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

  9. RISK-INFORMED SAFETY MARGIN CHARACTERIZATION

    International Nuclear Information System (INIS)

    Dinh, Nam; Szilard, Ronaldo

    2009-01-01

    The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system's 'loading' and its 'capacity', plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons

  10. Disclosing Risk Information: Assessing the Benefits

    International Nuclear Information System (INIS)

    Beierle, Thomas C.

    2003-01-01

    This paper describes a framework for examining the benefits of risk information disclosure and illustrates the framework through brief case studies of three information disclosure programs in the United States. I describe a general framework for analyzing the benefits of information disclosure and illustrate the framework by analyzing three disclosure programs in the United States: risk management planning (RMP), which provides detailed information on chemical accident risks and prevention; materials accounting, which provides information on how chemicals travel through processes at industrial facilities; and the Sector Facility Indexing Project (SFIP), which consolidates enforcement, compliance, and other data into a package of environmental performance indicators. This paper has outlined the types of benefits information disclosure programs ought to seek to achieve. Normative right-to-know benefits have been limited in some program because of intentionally circumscribed information sharing and apparent public disinterest. Substantive benefits have been more apparent, with many firms, agencies, NGOs, and others being able to point to the value of newly revealed information in better understanding environmental problems and the means to correct them. Instrumental benefits have been mixed, and firms appear to be responding to many motivations - not just public pressure - in deciding whether to improve environmental performance

  11. Forming homogeneous clusters for differential risk information

    International Nuclear Information System (INIS)

    Maardberg, B.

    1996-01-01

    Latent risk situations are always present in society. General information on these risk situations is supposed to be received differently by different groups of people in the population. In the aftermath of specific accidents different groups presumably have need of specific information about how to act to survive, to avoid injuries, to find more information, to obtain facts about the accidents etc. As targets for information these different groups could be defined in different ways. The conventional way is to divide the population according to demographic variables, such as age, sex, occupation etc. Another way would be to structure the population according to dependent variables measured in different studies. They may concern risk perception, emotional reactions, specific technical knowledge of the accidents, and belief in the information sources. One procedure for forming such groupings of people into homogeneous clusters would be by statistical clustering methods on dependent variables. Examples of such clustering procedures are presented and discussed. Data are from a Norwegian study on the perception of radiation from nuclear accidents and other radiation sources. Speculations are made on different risk information strategies. Elements of a research programme are proposed. (author)

  12. 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.

  13. 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

  14. 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.

  15. 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)

  16. 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.

  17. Inspection planning

    International Nuclear Information System (INIS)

    Korosec, D.; Levstek, M.F.

    2001-01-01

    Slovenian Nuclear Safety Administration (SNSA) division of nuclear and radiological safety inspection has developed systematic approach to their inspections. To be efficient in their efforts regarding regular and other types of inspections, in past years, the inspection plan has been developed. It is yearly based and organized on a such systematic way, that all areas of nuclear safety important activities of the licensee are covered. The inspection plan assures appropriate preparation for conducting the inspections, allows the overview of the progress regarding the areas to be covered during the year. Depending on the licensee activities and nature of facility (nuclear power plant, research reactor, radioactive waste storage, others), the plan has different levels of intensity of inspections and also their frequency. One of the basic approaches of the plan is to cover all nuclear and radiological important activities on such way, that all regulatory requests are fulfilled. In addition, the inspection plan is a good tool to improve inspection effectiveness based on previous experience and allows to have the oversight of the current status of fulfillment of planned inspections. Future improvement of the plan is necessary in the light of newest achievements on this field in the nuclear world, that means, new types of inspections are planned and will be incorporated into plan in next year.(author)

  18. 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...

  19. Risk-informed ranking of engineering projects

    International Nuclear Information System (INIS)

    Jyrkama, M.; Pandey, M.

    2011-01-01

    Refurbishment planning requires prudent investment decisions with respect to the various systems and components at the station. These decisions are influenced by many factors, including engineering, safety, regulatory, economic, and political constraints. From an engineering perspective, the concept of cost-benefit analysis is a common way to allocate capital among various projects. Naturally, the 'best' or optimal project should have the lowest cost and the highest benefit. In the context of risk-informed decision making (RIDM), a process that has been widely embraced by the global nuclear community, the costs and benefits must further be 'weighted' by probabilities to estimate the underlying risk associated with the various planning alternatives. The main purpose of this study is to illustrate how risk and reliability information can be integrated into the refurbishment planning process to facilitate more objective and transparent investment decisions. The methodology is based on the concept of generation risk assessment (GRA) which provides a systematic approach for balancing investment costs with the reduction in overall financial risk. In addition to reliability predictions, the model provides estimates for the level of risk reduction associated with each system/project and also the break-even point for investment. This information is vital for project ranking, and helps to address the key question of whether capital investment should be made in the most risk critical systems, or in systems that reduce the overall risk the most. The application of the proposed methodology requires only basic information regarding the current reliability of each engineering system, which should be readily available from plant records and routine condition assessments. Because the methodology can be readily implemented in a Microsoft Excel spreadsheet, all plausible (e.g., bounding) planning scenarios, with or without investment, can also be generated quickly and easily, while

  20. Inspection vehicle

    International Nuclear Information System (INIS)

    Takahashi, Masaki; Omote, Tatsuyuki; Yoneya, Yutaka; Tanaka, Keiji; Waki, Tetsuro; Yoshida, Tomiji; Kido, Tsuyoshi.

    1993-01-01

    An inspection vehicle comprises a small-sized battery directly connected with a power motor or a direct power source from trolly lines and a switching circuit operated by external signals. The switch judges advance or retreat by two kinds of signals and the inspection vehicle is recovered by self-running. In order to recover the abnormally stopped inspection vehicle to the targeted place, the inspection vehicle is made in a free-running state by using a clutch mechanism and is pushed by an other vehicle. (T.M.)

  1. Enhance pump reliability through improved inservice testing

    International Nuclear Information System (INIS)

    Healy, J.J.

    1990-01-01

    EPRI has undertaken a study to assess the effectiveness of existing testing programs to accurately monitor and predict performance changes before either pump performance degrades or an actual failure occurs. Anticipated changes in inservice testing techniques are directed towards enhancing the validity of test data, ensuring its repeatability, and avoiding deterioration of the pump assembly. There is a new-found interest in test programs of all types that has occurred, in part, because of an increase in reported pump degradation and pump failure. Inservice testing of pumps, which has long been a basis for assuring operability, has apparently produced an opposite effect; namely, the appearance of a reduction in reliability

  2. Discussion about risk-informed regulations on the nuclear safety

    International Nuclear Information System (INIS)

    Gu Yeyi

    2008-01-01

    The article introduces the background and status quo of regulations on the nuclear safety in China, and points out the inadequacies existing with the current regulations. The author explains the risk-informed safety management concerning its development, status quo, and achievements made, in an attempt to make out the trend of improving regulations on the nuclear safety through risk-informed methods. Combining the U.S. development program of establishing risk-informed regulations on the nuclear safety, the author narrates principles and features of the new regulations system, and provides suggestions for the promotion of risk-informed safety management and establishment of risk-informed regulations on the nuclear safety. (author)

  3. 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.)

  4. Ultrasonic inspection

    International Nuclear Information System (INIS)

    Satittada, Gannaga

    1984-01-01

    Ultrasonic inspection is one of the most widely used methods for nondestructive inspection. The beam of high-frequency sound wave, ultrasonic wave, is introduced into the material. It travels through the material with some attendant loss of energy and can be reflected at interfaces. The reflected beam is detected and analyzed. Ultrasonic inspection is used to detect flaws in metal parts as well as in welded, brazed and bonded joints during research work and developing production and service. It is also used to detect and locate porosity, pipe, and flakes. In addition, it can be used for the measurement of metal thickness. Ultrasonic inspection is therefore used for quality control and material inspection in all major industries

  5. Current products and future plan of regulatory technology R and D for risk-informed regulation and applications

    International Nuclear Information System (INIS)

    Song, K. Y.; Lee, C. J.; Kim, W. S.; Jeong, D. W.; Kim, H. J.

    2002-01-01

    The first phase of a R and D project for risk-informed regulation (RIR) and applications (RIA) has been finished. Various results which would be useful for preparing domestic RIR system were accomplished, in areas of safety goals and general principles of RIR, which provide fundamental bases for establishment of RIR system as well as regulatory review guides, which ensure the quality for PSA. RIA guidelines for ISI, IST, MOV, Tech.-Sepc. also have been developed, implementing some pilot plant applications. As essential documents for actual RIR inspection, risk-informed inspection guides and implementation guide for maintenance effectiveness were prepared. In the second phase of R and D, two projects on RIR area will be performed. One is to study on institutionalization of RIR and performance-based regulation, another is to develop a PSA model for regulatory audit as well as regulatory technology for risk monitoring

  6. 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)

  7. 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)

  8. 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.)

  9. 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.

  10. 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)

  11. 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

  12. In-service diagnostics of main circulating circuit pipes of WWER nuclear power plants

    International Nuclear Information System (INIS)

    Svoboda, V.; Merta, J.; Merta, V.

    1982-01-01

    The application is discussed of the acoustic emission method for testing the integrity of the components of the main circulating circuit of the WWER 440 nuclear power plant. A description is given of the main circulating circuit and a stress analysis on the basis of strength computations considering operating modes is presented. An analysis is also presented of the possible damage of the pipe material as related to the application of the acoustic emission method for in-service inspection of the pipes. Certain practical problems of application are discussed. (author)

  13. Hospital Inspections

    Data.gov (United States)

    U.S. Department of Health & Human Services — Welcome to hospitalinspections.org, a website run by the Association of Health Care Journalists (AHCJ) that aims to make federal hospital inspection reports easier...

  14. Outlook for risk assessment and risk-informed regulations

    International Nuclear Information System (INIS)

    Chakraborty, S.; Breutel, C.; Khatib-Rahbar, M.

    2000-01-01

    The Western nuclear regulatory process has evolved from the initial engineering judgment' framework of the 1960's, the prescriptive deterministic requirements of the 1970s, the transition years of the 1980s, to the present day movement toward risk-informed approaches. In this paper, a short overview of the historic development of safety regulation is provided. The critique of traditional regulatory practice will be summarized and the features of risk-informed regulation will be discussed. The implementation of risk-informed regulation is considered on the basis of general legal principles common to many member states of the NEA and IAEA. In a process to risk-inform regulation, principles such as equal treatment, proportionality of rules and predictability of administrative action are found to be important. (author)

  15. Nuclear Plant Modification in a Risk-Informed Environment

    International Nuclear Information System (INIS)

    Gallucci, Raymond H.V.

    2002-01-01

    This paper examines a specific nuclear power plant modification performed in a risk-informed regulatory environment. It quantifies both the permanent and temporary effects of the modification, and performs a cost-benefit evaluation. (authors)

  16. An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

    International Nuclear Information System (INIS)

    Ting, K.; Chen, K.T.; Li, Y.C.; Hwang, S.H.; Chien, F.T.; Kang, J.C.

    2008-01-01

    Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs' and AOVs' safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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.

  3. Operational inspections

    International Nuclear Information System (INIS)

    Bystersky, M.

    1997-01-01

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

  4. Use of risk information to safety regulation. Reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    A procedure of probabilistic risk assessment (PRA) for a reprocessing facility has been under the development aiming to utilize risk information for safety regulations in this project. Activities in the fiscal year 2012 are summarized in the paper. A major activity is a fundamental study on a concept of serious accidents, requirements of serious accident management, and a policy of utilizing risk information for fabrication and reprocessing facilities. Other than the activity a study on release and transport of aerial radioactive materials at a serious accident in a reprocessing facility has been conducted. The outline and results are provided in the chapter 1 and 2 respectively. (author)

  5. Use of risk information to safety regulation. Fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    A procedure of ISA (Integrated Safety Analysis) for uranium fuel fabrication/enrichment facilities has been under the development aiming to utilize risk information for safety regulations in this project. Activities in the fiscal year 2012 are summarized in the paper. There are two major activities in the year. First one is a study on ISA procedure for external events such as earthquakes. Second one is that for chemical consequences such as UF6 and HF. Other than the activities a fundamental study on a policy of utilizing risk information was conducted. The outline and results are provided in the chapter 1 and 2 respectively. (author)

  6. Regulatory use of risk information - initial developments at Slovenian Nuclear Safety Administration

    International Nuclear Information System (INIS)

    Muehleisen, A.; Koncar, M.; Vojnovic, D.; Persic, A.

    2004-01-01

    Similarly to other regulators worldwide, the SNSA intends to enhance the use of PSA and risk insights in its activities in order to ensure a better and more focused regulatory oversight as well as improved interface with a licensee. The main aim of the SNSA is to establish PSA as a standard tool to complement the deterministic based regulation for a variety of regulatory tasks. The PSA applications should, in particular, support the decision making process as well as the interactions with the Krsko NPP. As a first step in the internal use of PSA, PSA event analysis and risk based performance indicators are being introduced. In 2004, the SNSA will start introducing risk follow up and risk informed inspections. By mid 2005 the legal basis for the use of PSA will be also established in Slovenian legislation. (author)

  7. 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

  8. Supporting risk-informed decisions during business process execution

    NARCIS (Netherlands)

    Conforti, R.; Leoni, de M.; La Rosa, M.; Aalst, van der W.M.P.; Salinesi, C.; Norrie, M.C.; Pastor, O.

    2013-01-01

    This paper proposes a technique that supports process participants in making risk-informed decisions, with the aim to reduce the process risks. Risk reduction involves decreasing the likelihood and severity of a process fault from occurring. Given a process exposed to risks, e.g. a financial process

  9. Risk-informed approach in US-APWR technical specifications

    International Nuclear Information System (INIS)

    Saji, Etsuro; Tanaka, Futoshi; Kuroiwa, Katsuya; Kawai, Katsunori

    2009-01-01

    The Risk-Managed Technical Specifications and the Surveillance Frequency Control Program have been adopted in the US-APWR Technical Specifications. These risk-informed approaches are unique among the technical specifications for the advanced light water reactor designs adopted by planned nuclear power stations in the United States. (author)

  10. Risk Informed Design as Part of the Systems Engineering Process

    Science.gov (United States)

    Deckert, George

    2010-01-01

    This slide presentation reviews the importance of Risk Informed Design (RID) as an important feature of the systems engineering process. RID is based on the principle that risk is a design commodity such as mass, volume, cost or power. It also reviews Probabilistic Risk Assessment (PRA) as it is used in the product life cycle in the development of NASA's Constellation Program.

  11. Progress in EPRI-programs on the inspection of cast austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Dau, G; Behravesh, M; Amirato, P; Stone, R [Electric Power Research Inst., Charlotte, NC (United States). Nondestructive Evaluation Center

    1988-12-31

    This document presents the progress in EPRI programs on in-service inspection of Cast austenitic Stainless Steel (CSS). The CSS examination strategy is presented, together with results concerning thermal fatigue cracks and mechanical fatigue cracks. A statistical analysis method is provided, in order to estimate the crack detectability and the false call (a non-crack called crack). (TEC).

  12. 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…

  13. Technology Integration Support Levels for In-Service Teachers

    Science.gov (United States)

    Williams, Mable Evans

    2017-01-01

    In-service teachers across the globe are expected to integrate technology in their respective instructional content area. The purpose of this qualitative study was to explore the perceptions of in-service teachers concerning building-level support for technology integration. Participants in the study were asked to participate in semi-structured…

  14. 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…

  15. Inservice Teacher Education in Nigeria: A Case Study.

    Science.gov (United States)

    Esu, Akon E. O.

    1991-01-01

    Examines the current status of in-service teacher education in Nigeria, indicating three approaches: the central office approach; the long vacation program; and the Associateship Certificate in Education distance learning approach. Recommendations for planning and implementing in-service teacher education programs in Nigeria are noted. (SM)

  16. Institutional Roles for In-Service Education of School Administrators.

    Science.gov (United States)

    Lynch, Patrick D., Ed.; Blackstone, Peggy L., Ed.

    This document is a compilation of papers read at a 4-day conference attended by 60 participants from throughout the United States. Chapters include (1) "In-Service Education of School Administrators: Background, Present Status, and Problems," by Robert B. Howsam; (2) "Notes on Institutional Relationships in the In-Service Education of the…

  17. In-Service Teacher Education: Some Suggestions for Improvement ...

    African Journals Online (AJOL)

    This paper therefore examines teacher education, especially in-service teacher education and how it has been practised elsewhere. It is hoped that education policy makers will take note of some of the issues raised in this paper as the one day workshop which has hitherto been the most used strategy of in-service teacher ...

  18. A Leadership Model for University Geology Department Teacher Inservice Programs.

    Science.gov (United States)

    Sheldon, Daniel S.; And Others

    1983-01-01

    Provides geology departments and science educators with a leadership model for developing earth science inservice programs. Model emphasizes cooperation/coordination among departments, science educators, and curriculum specialists at local/intermediate/state levels. Includes rationale for inservice programs and geology department involvement in…

  19. Efficient improvement of nuclear power plant safety by reorganization of risk-informed safety importance evaluation methods for piping welded portions

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-09-01

    In this work, risk information was used to evaluate the safety importance of piping welded portions which were important for plant operation and maintenance of nuclear power plants. There are two types of risk-informed safety importance evaluation methods, namely the ASME method and the EPRI method. Since both methods have advantages and disadvantages, elements of each method were combined and reorganized. Considerations included whether the degradation mechanisms would be objectively evaluated and whether plant safety would be efficiently improved. The most objective and efficient method was as follows. Piping failure potential is quantitatively and objectively evaluated for failure with probabilistic fracture mechanics (PFM) and for other degradation mechanisms with empirical failure rates, and conditional core damage probability (CCDP) is calculated with PSA. This method reduces the inspected segment numbers to 1/4 of the deterministic method and increases the ratio of risk, which is covered by the inspected segments, to total risk from 80% of the deterministic method to 95%. Piping inspection numbers decreased for safety injection systems that were required the inspections by the deterministic method. Piping inspections were required for part of main feed water and main steam systems that were not required the inspections by the deterministic method. (author)

  20. 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

  1. 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)

  2. 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.

  3. Proposed rulemaking to risk-inform special treatment requirements

    International Nuclear Information System (INIS)

    Reed, Timothy A.; McKenna, Eileen M.

    2003-01-01

    This paper presents the status of Nuclear Regulatory Commission (NRC) rulemaking efforts to risk-inform special treatment requirements that reside in Title 10 of the Code of Federal Regulations, 10 CFR Part 21, Part 50, and Part 100. The staff has prepared a proposed rulemaking to add a new section to 10 CFR Part 50 to provide an alternative set of requirements for treatment of structures, systems and components (SSCs), using a risk-informed categorization process to determine safety significance of the SSCs. These requirements can be voluntarily adopted by light-water reactor licensees and applicants. The proposed rule is based upon extensive interactions with stakeholders (including consideration of public comments on draft rule language made available on the NRC rulemaking web site), experience with pilot plants, and guidance development activities. The NRC staff expects that stakeholder input provided in response to the proposed rule issuance will be valuable and support the efforts to issue the final rule. (author)

  4. Needs for Risk Informing Environmental Cleanup Decision Making - 13613

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Ming; Moorer, Richard [U.S. Department of Energy, Washington, DC 20585 (United States)

    2013-07-01

    This paper discusses the needs for risk informing decision making by the U.S. Department of Energy (DOE) Office of Environmental Management (EM). The mission of the DOE EM is to complete the safe cleanup of the environmental legacy brought about from the nation's five decades of nuclear weapons development and production and nuclear energy research. This work represents some of the most technically challenging and complex cleanup efforts in the world and is projected to require the investment of billions of dollars and several decades to complete. Quantitative assessments of health and environmental risks play an important role in work prioritization and cleanup decisions of these challenging environmental cleanup and closure projects. The risk assessments often involve evaluation of performance of integrated engineered barriers and natural systems over a period of hundreds to thousands of years, when subject to complex geo-environmental transformation processes resulting from remediation and disposal actions. The requirement of resource investments for the cleanup efforts and the associated technical challenges have subjected the EM program to continuous scrutiny by oversight entities. Recent DOE reviews recommended application of a risk-informed approach throughout the EM complex for improved targeting of resources. The idea behind this recommendation is that by using risk-informed approaches to prioritize work scope, the available resources can be best utilized to reduce environmental and health risks across the EM complex, while maintaining the momentum of the overall EM cleanup program at a sustainable level. In response to these recommendations, EM is re-examining its work portfolio and key decision making with risk insights for the major sites. This paper summarizes the review findings and recommendations from the DOE internal reviews, discusses the needs for risk informing the EM portfolio and makes an attempt to identify topics for R and D in

  5. Needs for Risk Informing Environmental Cleanup Decision Making - 13613

    International Nuclear Information System (INIS)

    Zhu, Ming; Moorer, Richard

    2013-01-01

    This paper discusses the needs for risk informing decision making by the U.S. Department of Energy (DOE) Office of Environmental Management (EM). The mission of the DOE EM is to complete the safe cleanup of the environmental legacy brought about from the nation's five decades of nuclear weapons development and production and nuclear energy research. This work represents some of the most technically challenging and complex cleanup efforts in the world and is projected to require the investment of billions of dollars and several decades to complete. Quantitative assessments of health and environmental risks play an important role in work prioritization and cleanup decisions of these challenging environmental cleanup and closure projects. The risk assessments often involve evaluation of performance of integrated engineered barriers and natural systems over a period of hundreds to thousands of years, when subject to complex geo-environmental transformation processes resulting from remediation and disposal actions. The requirement of resource investments for the cleanup efforts and the associated technical challenges have subjected the EM program to continuous scrutiny by oversight entities. Recent DOE reviews recommended application of a risk-informed approach throughout the EM complex for improved targeting of resources. The idea behind this recommendation is that by using risk-informed approaches to prioritize work scope, the available resources can be best utilized to reduce environmental and health risks across the EM complex, while maintaining the momentum of the overall EM cleanup program at a sustainable level. In response to these recommendations, EM is re-examining its work portfolio and key decision making with risk insights for the major sites. This paper summarizes the review findings and recommendations from the DOE internal reviews, discusses the needs for risk informing the EM portfolio and makes an attempt to identify topics for R and D in integrated

  6. Risk-Informed Assessment Methodology Development and Application

    International Nuclear Information System (INIS)

    Sung Goo Chi; Seok Jeong Park; Chul Jin Choi; Ritterbusch, S.E.; Jacob, M.C.

    2002-01-01

    Westinghouse Electric Company (WEC) has been working with Korea Power Engineering Company (KOPEC) on a US Department of Energy (DOE) sponsored Nuclear Energy Research Initiative (NERI) project through a collaborative agreement established for the domestic NERI program. The project deals with Risk-Informed Assessment (RIA) of regulatory and design requirements of future nuclear power plants. An objective of the RIA project is to develop a risk-informed design process, which focuses on identifying and incorporating advanced features into future nuclear power plants (NPPs) that would meet risk goals in a cost-effective manner. The RIA design methodology is proposed to accomplish this objective. This paper discusses the development of this methodology and demonstrates its application in the design of plant systems for future NPPs. Advanced conceptual plant systems consisting of an advanced Emergency Core Cooling System (ECCS) and Emergency Feedwater System (EFWS) for a NPP were developed and the risk-informed design process was exercised to demonstrate the viability and feasibility of the RIA design methodology. Best estimate Loss-of-Coolant Accident (LOCA) analyses were performed to validate the PSA success criteria for the NPP. The results of the analyses show that the PSA success criteria can be met using the advanced conceptual systems and that the RIA design methodology is a viable and appropriate means of designing key features of risk-significant NPP systems. (authors)

  7. Development of a risk informed fire protection program

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, J.; McDevitt, B.; Sawyer, O.; Volk, M.A.; Drennan, J.; Sweely, C.

    2015-07-01

    Over the past decade, one of the largest challenges for the Nuclear Power Plant (NPP) Operator in the United States has been the implementation of risk-informed, performance-based (RI-PB) fire protection strategies into their fire protection program. Regardless of whether a utility decides to fully transition their licensing basis from deterministic to risk based, or if they simply complete a fire probabilistic risk assessment (FPRA) in order to augment their current program, it is clear that risk-informed, performance based fire protection strategies and the associated challenges are the growing trend in the United States and are here to stay. The experience of the nuclear industry in the United States with the implementation of RI-PB fire protection strategies can provide a great deal of insight for plants and utilities that follow, either by choice or necessity, a similar path. The similarities in the design of the United States and Spanish nuclear plants make these insights even more significant contributions to the strategy and planning for the Spanish fleet. The experience in United States will provide guidance to avoid similar missteps and better plan for the challenges of the transition process. As the Spanish fleet develops risk-informed and deterministic strategies to improve fire safety, an understanding of the challenges and lessons learned from the United States experience will save time and money. (Author)

  8. Risk-informed design of a pebble bed gas reactor

    International Nuclear Information System (INIS)

    Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

    2003-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

  9. A perspective on risk informed regulation and the maintenance rule

    International Nuclear Information System (INIS)

    In, Young H.

    2006-01-01

    Establishing a new regulatory framework for the future can be a complex and difficult task where the current regulatory framework and the past regulatory history must be reconciled. The nuclear regulatory framework will evolve over time, and it is in our best interest to make the right changes, whether the change is incremental or innovative. Basically, the risk-informed regulation (RIR) is an innovative change that must be made incrementally. Therefore, it is important that the changes to the framework are carefully evaluated and studied before the actual implementation. Otherwise, we may see unnecessary patches along the way that can be damaging to the regulatory integrity and the public confidence in the regulation. Even though, the United States has led the risk-informed regulation implementation during last decade, there were some instances where the well intended changes were not well planned and coordinated, resulting in loss of valuable time and resources that led to confusions among involved parties. The risk-informed regulation is an emerging trend worldwide, and Korea is one of the few that are ahead of the pack in embracing this innovative change in the regulation. To better understand and plan for RIR, we need to study both the well-established regulatory framework and the emerging regulatory framework. In addition, the recent trends in other regulatory framework similar to its own should be carefully studied. The reasons for any stagnations or barriers in other regulatory frameworks should be carefully noted and see if we share similar problems so that we can formulate better solutions. Blindly following others regulatory changes is not recommended, since our regulatory environment is unique in itself. We should take their inputs and lessons learned, but adopt them specifically to our industry needs and regulatory environment. This paper will attempt to provide a perspective on the risk-informing process of a nuclear plant by a brief look into two key

  10. Risk informed regulation of nuclear facilities: Overview of the current status

    International Nuclear Information System (INIS)

    2005-02-01

    This report provides guidance on the use of risk information by regulatory bodies as part of an integrated decision making process. This addresses the way in which risk information is being used in decisions about safety issues at nuclear plants, sometimes referred to as risk informed decision making, and how risk information is being used by regulatory bodies as an input into the activities that they carry out, sometimes referred to as risk informed regulation

  11. 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

  12. 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

  13. 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

  14. A historical perspective of risk-informed regulation

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, P.L.

    1996-12-01

    In Federal studies, the process of using risk information is described as having two general components: (1) risk assessment - the application of credible scientific principles and statistical methods to develop estimates of the likely effects of natural phenomena and human factors and the characterization of these estimates in a form appropriate for the intended audience (e.g., agency decisionmakers, public); and (2) risk management - the process of weighing policy alternatives and selecting the most appropriate regulatory action, integrating the results of risk assessment with engineering data with social, economic, and political concerns to reach a decision. This paper discusses largely the second component.

  15. A historical perspective of risk-informed regulation

    International Nuclear Information System (INIS)

    Campbell, P.L.

    1996-01-01

    In Federal studies, the process of using risk information is described as having two general components: (1) risk assessment - the application of credible scientific principles and statistical methods to develop estimates of the likely effects of natural phenomena and human factors and the characterization of these estimates in a form appropriate for the intended audience (e.g., agency decisionmakers, public); and (2) risk management - the process of weighing policy alternatives and selecting the most appropriate regulatory action, integrating the results of risk assessment with engineering data with social, economic, and political concerns to reach a decision. This paper discusses largely the second component

  16. Risk-Informed SSCs Categorization: Elicitation Method of Expert's Opinion

    International Nuclear Information System (INIS)

    Hwang, Mee Jeong; Yang, Joon Eon; Kim, Kil Yoo

    2005-01-01

    The regulations have been performing by deterministic way since nuclear power plants have been operating. However, some SSCs identified as safety-significance by deterministic way, were turned out to be low or non safety-significant and some SSCs identified as non-safety significance were turned out to be high safety-significant according to the results of PSA. Considering these risk insights, Regulatory Guide 1.174 and 10CFR50.69 were drawn up, and we can re-categorize the SSCs according to their safety significance. Therefore, a study and an interest about the risk-informed SSCs re-categorization and treatment has been continued. The objective of this regulatory initiative is to adjust the scope of equipment subject to special regulatory treatment to better focus licensee and regulatory attention and resources on equipment that has safety significance. Current most regulations define the plant equipment necessary to meet deterministic regulatory basis as 'safety-related.' This equipment is subject to special treatment regulations. Other plant equipment is categorized as 'non-safety related,' and is not subject to a select number of special treatment requirement or a subset of those requirement. However, risk information is not a magic tool making a decision but a supporting tool to categorize SSCs. This is because only small parts of a plant are modeled in PSA model. Thus, engineering and deterministic judgments are also used for risk-informed SSCs categorization, and expert opinion elicitation is very important for risk-informed SSCs categorization. Therefore, we need a rational method to elicit the expert's opinions, and in this study, we developed a systematic method for expert elicitation to categorize the nuclear power plants' SSCs. Current states for SSCs categorization of the USA and the existing methods for expert elicitation were surveyed and more systematic way eliciting the expert opinions and combining was developed. To validate the developed method

  17. 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

  18. 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

  19. 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)

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

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

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

  2. Risk-informed decision making during Bohunice NPP safety upgrading

    International Nuclear Information System (INIS)

    Lipar, M.; Muzikova, E.; Kubanyi, J.

    2001-01-01

    The paper summarizes some facts of risk-informed regulation developments within UJD regulatory environment. Based on national as well as international operating experience and indications resulted from PSA, Nuclear Regulatory Authority of the Slovak Republic (UJD) since its constituting in 1993 has devoted an effort to use PSA technology to support the regulatory policy in Slovakia. The PSA is considered a complement, not a substitute, to the deterministic approach. Suchlike integrated approach is used in decision making processes and the final decision on scope and priorities is based on it. The paper outlines risk insights used in the decision making process concerning Bohunice NPP safety upgrading and focuses on the role of PSA results in Gradual Reconstruction of Bohunice VI NPP. Besides, two other examples of the PSA results application to the decision making process are provided: the assessment of proposal of modifications to the main power supply diagram (incorporation of generator switches) and the assessment of licensee request for motor generator AOT (Allowable Outage Time) extension. As an example of improving support of Bohunice V-2 risk-informed operations, concept of AOT calculations and Bohunice V-2 Risk Monitor Project are briefly described. (author)

  3. Risk informed decision making. Topical issues paper no. 1

    International Nuclear Information System (INIS)

    Niehaus, F.; Szikszai, T.

    2001-01-01

    To date, probabilistic safety assessments (PSAs) have been performed for more than 200 nuclear power plants (NPPs) worldwide and are under various stages of development for most of the remaining NPPs. The state-of-the-art is to have a full scope Level 2 PSA (including external events and low power and shutdown) which is maintained as a 'living PSA' with regular updating. Modern computer technology allows frequent recalculations of the PSA to evaluate the impact of changes in operation or design and allows use of the PSA in the form of safety or risk monitors. There is a general agreement, as documented in various IAEA Safety Standards, that the deterministic approach to nuclear safety should be complemented by a probabilistic approach. Though PSAs have been used extensively in the past, it was usually limited to a variety of applications on a case by case basis as deemed necessary or useful. There is now a recent development led by the USA, and followed by several other countries, to move to a much expanded use of PSA in what is termed 'risk informed decision making'. The main driving force behind this movement is the expectation that the use of risk insights can result in both improved safety and a reduction in unnecessary regulatory requirements, hence leading to a more efficient use of resources for NPP operators and the regulatory authority. One of the key challenges in truly risk informed decision making is the reconciliation of PSA results and insights with traditional deterministic analysis. This is particularly true when it comes to defence in depth and safety margins. PSA results often conflict with deterministic insights. If a method of reconciling these conflicts is not defined, then risk informed can become deterministic plus PSA. This results in PSA being an additional layer of requirements rather than a tool for optimized decision making. Alternatively, if PSA information is always used to override deterministic considerations, then that is a 'risk

  4. In-situ inspection of grooves in reactor tube sheet using a remotely operated cast impression taking device

    International Nuclear Information System (INIS)

    Rajendran, S.; Ramakumar, M.S.

    1996-01-01

    Utmost importance is given to the in-service inspection of critical components of a reactor to ensure its reliable performance during the reactor operation. This paper describes a cast taking device using cold setting resin to take impression of the grooves being made in the tube sheet for sparger tube installation in pressurised heavy water reactor. (author)

  5. 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.)

  6. A quantitative approach for risk-informed safety significance categorization in option-2

    International Nuclear Information System (INIS)

    Ha, Jun Su; Seong, Poong Hyun

    2004-01-01

    OPTION-2 recommends that Structures, Systems, or Components (SSCs) of Nuclear Power Plants (NPPs) should be categorized into four groups according to their safety significance as well as whether they are safety-related or not. With changes to the scope of SSCs covered by 10 CFR 50, safety-related components which categorized into low safety significant SSC (RISC-3 SSC) can be exempted from the existing conservative burden (or requirements). As OPTION-2 paradigm is applied, a lot of SSCs may be categorized into RISC-3 SSCs. Changes in treatment of the RISC-3 SSCs will be recommended and then finally the recommended changes shall be evaluated. Consequently, before recommending the changes in treatment, probable candidate SSCs for the changes in treatment need to be identified for efficient risk-informed regulation and application (RIRA). Hence, in this work, a validation focused on the RISC-3 SSCs is proposed to identify probable candidate SSCs. Burden to Importance Ratio (BIR) is utilized as a quantitative measure for the validation. BIR is a measure representing the extent of resources or requirements imposed on a SSC with respect to the value of the importance measure of the SSC. Therefore SSCs having high BIR can be considered as probable candidate SSCs for the changes in treatment. In addition, the final decision whether RISC-3 SSCs can be considered as probable candidate SSCs or not should be made by an expert panel. For the effective decision making, a structured mathematical decision-making process is constructed based on Belief Networks (BBN) to overcome demerits of conventional group meeting based on unstructured discussion for decision-making. To demonstrate the usefulness of the proposed approach, the approach is applied to 22 components selected from 512 In-Service Test (IST) components of Ulchin unit 3. The results of the application show that the proposed approach can identify probable candidate SSCs for changes in treatment. The identification of the

  7. Status on system inspection and preventive maintenance of HANARO

    International Nuclear Information System (INIS)

    Kim, Young-Ki; Cho, Yeong-Garp; Kwag, Byung-Ho

    1999-01-01

    The HANARO is a 30 MW open pool type multi-purpose research reactor with forced light water coolant/moderator flows and heavy water annular reflector. The relatively small reactor core uses a low enriched fuel and is designed to maximize the power density, thus providing the required neutron flux for various research activities. It is mainly used for radioisotope production, nuclear material testing and neutron physics experiments. The initial criticality was achieved February 1995. Considering the importance of their functionality from the safety point of view, some components and equipment are categorized into a nuclear safety grade. There are three different inspection activities for the various reactor systems and components - a Surveillance Inspection(SI) for the safety grades and a Periodic Inspection (PI) for the non-safety grades and In-Service Inspection (ISI) for the ASME Sec.III components. All of the SIs are specified and required by the safety analysis report. The SI also differs from the PI in such a point that all kinds of activities for the SIs should be accompanied by an appropriate quality assurance, while for the PIs it is not necessarily mandatory. In addition, the inspection results for the SIs should go through an examination from regulatory body every two years and specific functions of the critical components or systems are demonstrated under the witness by the governmental inspector. The ISI is required and carried out as per international codes and standards as well as Korean atomic energy regulations. There are 54 SIs, 25 PIs and 4 ISIs for the HANARO. This paper concentrates on the managing strategy and its practices for the SIs and ISIs of the safety-related components, currently being done at HANARO. Most parts of the inspections fall into a group for the periodic performance testing and/or equipment calibration. Some mechanical inspections like a torque measurement are grouped into a preventive maintenance. Lastly the ASME Sec

  8. Risk-informed assessment of degraded containment structures

    International Nuclear Information System (INIS)

    Spencer, B.W.; Kunsman, D.M.; Graves, H.L.

    2003-01-01

    As nuclear power plants age, a number of degradation mechanisms may begin to affect the ability of critical containment structures to prevent radiation release during a severe accident. A research program is underway to quantify the effects of various types of containment degradation in a risk-informed manner. In this paper, corrosion is assumed to occur in the liner of a reinforced concrete containment at a 'typical' U.S. pressurized water reactor nuclear power plant, and its effect is investigated. Latin hypercube sampling is used in conjunction with finite element models of a typical steel-lined reinforced concrete containment to generate overpressurization fragilities of the containment with and without corrosion. An existing probabilistic risk assessment model of the plant is then used with these fragilities to determine the increase in risk caused by the corrosion. (author)

  9. Development of risk-informed assessment (RIA) design methodology

    International Nuclear Information System (INIS)

    Ji, S. K.; Park, S. J.; Park, B. R.; Kim, M. R.; Choi, C. J.

    2001-01-01

    It has been assessed that the capital cost for future nuclear power plants needs to be reduced on the order of 35% to 40% for Advanced Light Water Reactors such as KNGR and System 80+. Such reduction in the capital cost will require a fundamental re-evaluation of the industry standards and regulatory basis under which nuclear plants are designed and licensed. The objective of this study is to develop the risk-informed assessment (RIA) design methodology for future nuclear power plants. In order to meet this objective, the design simplification method is developed and RIA design methodology exercised for conceptual system. For the methodology verification, simplified conceptual ECCS and feedwater system are developed, then LOCA sensitivity analyses and agressive secondary cooldown analyses for these systems are performed. In addition, the probability safety assessment (PSA) model for LOCA is developed and the validation of RIA design methodology is demonstrated

  10. Use of importance measures in risk-informed regulatory applications

    International Nuclear Information System (INIS)

    Cheok, Michael C.; Parry, Gareth W.; Sherry, Richard R.

    1998-01-01

    The use of importance measures to analyze PRA results is discussed. Commonly used importance measures are defined. Some issues that have been identified as potentially limiting their usefulness are addressed, namely: there is no simple relationship between importance measures evaluated at the single component level and those evaluated at the level of a group of components, and, as a result, some of the commonly used importance measures are not realistic measures of the sensitivity of the overall risk to parameter value changes; and, importance measures do not typically take into account parameter uncertainties which raises the question of the robustness of conclusions drawn from importance analyses. The issues are explored in the context of both ranking and categorization of structures, systems, and components (SSCs) with respect to risk-significance and safety-significance for use in risk-informed regulatory analyses

  11. Regulatory approach to risk informed decision making in India

    International Nuclear Information System (INIS)

    Chande, S.K.; Koley, J.

    2001-01-01

    Atomic Energy Regulatory Board (AERB), the authority for licensing and monitoring safety in Indian Nuclear Power Plants (NPPs), makes use of insights gained from PSA together with the results of the other deterministic analyses in taking decisions regarding the acceptability of the safety of the NPPs. PSA provides an estimation of risks; it also gives information on a balanced design by revealing interaction between engineered features and weak areas in a design. For regulatory use, PSA needs to be carried out using standardized methodology and state of the art technology. PSA helps regulators in taking faster and consistent decisions. Keeping in mind the limitations associated with PSA study, AERB has decided to adopt risk-informed decision making in regulatory licensing process. This paper describes the AERB policy regarding PSA and gives an overview of the experience in this area. (author)

  12. Risk-informing special treatment requirements for reactors

    International Nuclear Information System (INIS)

    McKenna, E.M.; Reed, T.A.

    2001-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is proposing to make regulatory changes to the scope of structures, systems, and components (SSCs) requiring special treatment. ''Special treatment requirements'' refers to those specific examples of regulations that are applied in order to provide a high degree of assurance that SSC will be capable of performing their intended functions when needed. The current scope of SSCs covered by the special treatment requirements governing commercial nuclear reactors is deterministically based and stems primarily from the evaluation of selected design basis events, as described in updated final safety analysis reports (UFSARs). This regulatory framework provides reasonable assurance of no undue risk to the health and safety of the public. However, recent advances in technology, coupled with operating reactor experience, have suggested that an alternative approach that would use a risk-informed process for evaluating SSC safety significance, would, in turn, result in a more focused determination of which SSCs should receive special treatment requirements. (author)

  13. In-service diagnostic systems of steam generators, pressurizers and other components of WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.

    1988-01-01

    A detailed description is presented of the systems of vibration inspections and systems of determining residual service life, implemented as in-service diagnostic systems for steam generators and pressurizers at the Dukovany nuclear power plant. Low temperature accelerometers of the KD or KS type and high temperature accelerometers CA 91 are used as vibration sensors. In the system of vibration inspection a total of 64 vibration measuring chains of Czechoslovak make and design are installed in the power plant. Systems are being built for determining residual service life which consist of 75 special chains for heat monitoring with thermocouples installed on selected assemblies of the steam generators and the pressurizers serving to monitor and evaluate heat stress. Also included in the system for determining residual service life are 16 routes for water withdrawal from steam generators. Their purpose is to make in-service determinations of places of biggest concentrations of impurities in secondary water, to determine the biggest local chemical exposure of primary collector and heat exchange tube materials and to optimize the size and place of leachate withdrawal. (Z.M.). 2 figs., 2 tabs., 15 refs

  14. Risk-informed, performance-based safety-security interface

    International Nuclear Information System (INIS)

    Mrowca, B.; Eltawila, F.

    2012-01-01

    Safety-security interface is a term that is used as part of the commercial nuclear power security framework to promote coordination of the many potentially adverse interactions between plant security and plant safety. Its object is to prevent the compromise of either. It is also used to describe the concept of building security into a plant's design similar to the long standing practices used for safety therefore reducing the complexity of the operational security while maintaining or enhancing overall security. With this in mind, the concept of safety-security interface, when fully implemented, can influence a plant's design, operation and maintenance. It brings the approach use for plant security to one that is similar to that used for safety. Also, as with safety, the application of risk-informed techniques to fully implement and integrate safety and security is important. Just as designers and operators have applied these techniques to enhance and focus safety, these same techniques can be applied to security to not only enhance and focus the security but also to aid in the implementation of effective techniques to address the safety-security interfaces. Implementing this safety-security concept early within the design process can prevent or reduce security vulnerabilities through low cost solutions that often become difficult and expensive to retrofit later in the design and/or post construction period. These security considerations address many of the same issues as safety in ensuring that the response of equipment and plant personnel are adequate. That is, both safety and security are focused on reaching safe shutdown and preventing radiological release. However, the initiation of challenges and the progression of actions in response these challenges and even the definitions of safe shutdown can be considerably different. This paper explores the techniques and limitations that are employed to fully implement a risk-informed, safety-security interface

  15. Improvement of infrastructure for risk-informed regulation

    International Nuclear Information System (INIS)

    2012-01-01

    Improvement of the infrastructure of probabilistic safety assessment (PSA) is essential to the risk-informed regulation for nuclear power plants. JNES conducted update of initiating event frequencies and improvement of the method for uncertainty analysis to enhance the technology bases of PSA in 2011. Furthermore, JNES improved the human reliability analysis method and the reliability analysis method for digital reactor protection systems. JNES estimated initiating event frequencies both for power and shutdown operation based on the recent operating experiences in Nuclear Power Plants (NPPs) of Japan using the hierarchical Bayesian method. As for improvement of the uncertainty analysis method, JNES conducted trial analyses using SOKC (State-Of-Knowledge Correlation) for the representative PWR plant and BWR plant of Japan. The study on the advanced HRA method with operator cognitive action model was conducted to improve a quality of HRA. The study on analyses of 'defense in depth' and 'diversity' for introducing digital instrumentation and control (I and C) systems was conducted. In order to ensure the quality of PSA, JNES conducted a peer review of a representative Japanese BWR plant PSA by the professional PSA engineers from the U.S. in order to extract to improve quality of PSA, and made an effort to develop the procedures of internal fire PSA. JNES participated in OECD/NEA PRISME and FIRE project to obtain the latest information and data to validate and improve the fire propagation analysis codes and the parameters for fire PSA as well. Furthermore, JNES studied schemes for the endorsement and application in the risk-informed regulation of PSA standards established by Atomic Energy Society of Japan. (author)

  16. Improvement of infrastructure for risk-informed regulation

    International Nuclear Information System (INIS)

    Muta, Hitoshi; Tanji, Junichi; Kondo, Keisuke; Uchida, Tsuyoshi; Ito, Tomomichi

    2011-01-01

    Improvement of the infrastructure of probabilistic safety assessment (PSA) is essential to the risk-informed regulation for nuclear power plants. JNES conducted update of initiating event frequency and improvement of method for uncertainty analysis to enhance the technology bases of PSA in 2010. Furthermore, JNES improved human reliability assessment method and reliability assessment method for digital reactor protection systems. JNES estimated initiating event frequencies both for power and shutdown operation based on the recent operating experiences in NPPs of Japan using hierarchical Bayesian method. As for improvement of uncertainty analysis method, JNES conducted trial analysis using SOKC (State-Of-Knowledge Correlation) for representative PWR and BWR of Japan. The study on the advanced HRA method with operator cognitive action model was conducted. The study on reliability analysis method for digital reactor protection systems using Bayesian Network Method was conducted. In order to ensure the quality of PSA, JNES studied requirements and methods for PSA peer review via the preparation of peer review for PSA of a representative Japanese BWR plant conducted by JNES. As an effort to develop the procedures of internal fire PSA and internal flooding PSA, trial analyses were conducted to grasp the risk level cause by fire and flooding in nuclear power plants. JNES participated in OECD/NEA PRISME and FIRE project to obtain the latest information and data to validate and improve the fire propagation analysis codes and the parameters for fire PSA. Furthermore, JNES studies schemes for endorsement and application in risk-informed regulation of PSA standards established by Atomic Energy Society of Japan. (author)

  17. Planning for risk-informed/performance-based fire protection at nuclear power plants. Final report

    International Nuclear Information System (INIS)

    Najafi, B.; Parkinson, W.J.; Lee, J.A.

    1997-12-01

    This document presents a framework for discussing issues and building consensus towards use of fire modeling and risk technology in nuclear power plant fire protection program implementation. The plan describes a three-phase approach: development of core technologies, implementation of methods, and finally, case studies and pilot applications to verify viability of such methods. The core technologies are defined as fire modeling, fire and system tests, use of operational data, and system and risk techniques. The implementation phase addresses the programmatic issues involved in implementing a risk-informed/performance-based approach in an integrated approach with risk/performance measures. The programmatic elements include: (1) a relationship with fire codes and standards development as defined by the ongoing effort of NFPA for development of performance-based standards; (2) the ability for NRC to undertake inspection and enforcement; and (3) the benefit to utilities in terms of cost versus safety. The case studies are intended to demonstrate applicability of single issue resolution while pilot applications are intended to check the applicability of the integrated program as a whole

  18. 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)

  19. 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

  20. Additive Manufacturing Infrared Inspection

    Science.gov (United States)

    Gaddy, Darrell; Nettles, Mindy

    2015-01-01

    The Additive Manufacturing Infrared Inspection Task started the development of a real-time dimensional inspection technique and digital quality record for the additive manufacturing process using infrared camera imaging and processing techniques. This project will benefit additive manufacturing by providing real-time inspection of internal geometry that is not currently possible and reduce the time and cost of additive manufactured parts with automated real-time dimensional inspections which deletes post-production inspections.

  1. The Preliminary Review for the Cross-Cutting Issues in the US Nuclear Regulatory Inspection Framework

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Suk; Jung, Dae Wook [Future and Challenges Inc., Seoul (Korea, Republic of); Cho, Nam Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-10-15

    The research for the development of risk-informed and performance-based regulatory inspection is ongoing in KINS. In the USNRC, the cross-cutting issue is one of the main components the risk-informed and performance-based regulatory inspection process as shown in figure 1, which is named as ROP (Reactor Oversight Process). The following three cross-cutting areas implicitly affect all of the safety cornerstones in ROP. In this study, the preliminary review for the inspection practices of cross-cutting issues in the US and Korean safety regulatory system were performed. The elements of the cross-cutting issues were recently modified to emphasize the importance of safety culture, and the graded approach was applied for the inspection of cross-cutting issues in USNRC. The graded approach for the inspection of cross-cutting issues will be also needed to Korean safety regulatory system in the future.

  2. The Preliminary Review for the Cross-Cutting Issues in the US Nuclear Regulatory Inspection Framework

    International Nuclear Information System (INIS)

    Lee, Yong Suk; Jung, Dae Wook; Cho, Nam Chul

    2008-01-01

    The research for the development of risk-informed and performance-based regulatory inspection is ongoing in KINS. In the USNRC, the cross-cutting issue is one of the main components the risk-informed and performance-based regulatory inspection process as shown in figure 1, which is named as ROP (Reactor Oversight Process). The following three cross-cutting areas implicitly affect all of the safety cornerstones in ROP. In this study, the preliminary review for the inspection practices of cross-cutting issues in the US and Korean safety regulatory system were performed. The elements of the cross-cutting issues were recently modified to emphasize the importance of safety culture, and the graded approach was applied for the inspection of cross-cutting issues in USNRC. The graded approach for the inspection of cross-cutting issues will be also needed to Korean safety regulatory system in the future

  3. A risk informed safety classification for a Nordic NPP

    International Nuclear Information System (INIS)

    Jaenkaelae, K.

    2002-01-01

    The report describes a study to develop a safety classification proposal or classi- fication recommendations based on risks for selected equipment of a nuclear power plant. The application plant in this work is Loviisa NPP unit 1. The safety classification proposals are to be considered as an exercise in this pilot study and do not necessarily represent final proposals in a real situation. Comparisons to original safety classifications and technical specifications were made. The study concludes that it is possible to change safety classes or safety signifi- cances as considered in technical specifications and in in-service-inspections into both directions without endangering the safety or even by improving the safety. (au)

  4. Risk Information Seeking among U.S. and Dutch Residents. An Application of the model of Risk Information Seeking and Processing

    NARCIS (Netherlands)

    ter Huurne, E.F.J.; Griffin, Robert J.; Gutteling, Jan M.

    2009-01-01

    The model of risk information seeking and processing (RISP) proposes characteristics of individuals that might predispose them to seek risk information. The intent of this study is to test the model’s robustness across two independent samples in different nations. Based on data from the United

  5. Change of risk information disclosure in annual report. Before and after earthquake disaster

    International Nuclear Information System (INIS)

    Ueno, Takefumi

    2011-01-01

    This research examines how risk information disclosure is changing in annual report before and after East Japan Great Earthquake Disaster. Company voluntary disclose risk information in annual report. Manager can decide a style and items of risk information. This paper explores risk information disclosures of Tokyo Electric Power Company, Chubu Power Electric Company, Kansai Electric Power Company and Toyota Motor Corporation. The managers except Tokyo Electric Company are likely to disclose own catastrophe risk before the disaster. However, they do not try to reduce their risk. Corporations' risk information do not link with own risk management. (author)

  6. Toward introduction of risk informed safety regulation. Nuclear Safety Commission taskforce's interim report

    International Nuclear Information System (INIS)

    2006-01-01

    Nuclear Safety Commission's taskforce on 'Introduction of Safety Regulation Utilizing Risk Information' completed the interim report on its future subjects and directions in December 2005. Although current safety regulatory activities have been based on deterministic approach, this report shows the risk informed approach is expected to be very useful for making nuclear safety regulation and assurance activities reasonable and also for appropriate allocation of regulatory resources. For introduction of risk informed regulation, it also recommends pileups of experiences with gradual introduction and trial of the risk informed approach, improvement of plant maintenance rules and regulatory requirements utilizing risk information, and establishment of framework to assure quality of risk evaluation. (T. Tanaka)

  7. Development of inspection and maintenance program for reactor and reactivity control units in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp

    1998-01-01

    This paper summarizes the overall program for inspection and maintenance of reactor structure and Reactivity Control Units (RCU) of HANARO during lifetime. The long-term plan for in-service inspection is introduced in the viewpoint of the structural integrity of reactor and RCU, and the operability of RCU mechanism. This program includes the list of components to be inspected, the schedule of inspection and maintenance, and the development of special tools and test rig that are required for the remote inspection and maintenance of reactor and RCU components. Preliminary results of the evaluation on the lifetime of RCU components are summarized based on the operation history since the installation of reactor. A test rig will be designed and constructed for the purposes of verifying the prolonged lifetime of RCU components being used, the performance of special tools, and the rehearsal of maintenance work as well. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  9. Characterization of Developer Application Methods Used in Fluorescent Penetrant Inspection

    Science.gov (United States)

    Brasche, L. J. H.; Lopez, R.; Eisenmann, D.

    2006-03-01

    Fluorescent penetrant inspection (FPI) is the most widely used inspection method for aviation components seeing use for production as well as an inservice inspection applications. FPI is a multiple step process requiring attention to the process parameters for each step in order to enable a successful inspection. A multiyear program is underway to evaluate the most important factors affecting the performance of FPI, to determine whether existing industry specifications adequately address control of the process parameters, and to provide the needed engineering data to the public domain. The final step prior to the inspection is the application of developer with typical aviation inspections involving the use of dry powder (form d) usually applied using either a pressure wand or dust storm chamber. Results from several typical dust storm chambers and wand applications have shown less than optimal performance. Measurements of indication brightness and recording of the UVA image, and in some cases, formal probability of detection (POD) studies were used to assess the developer application methods. Key conclusions and initial recommendations are provided.

  10. 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

  11. Which Type of Risk Information to Use for Whom? Moderating Role of Outcome-Relevant Involvement in the Effects of Statistical and Exemplified Risk Information on Risk Perceptions.

    Science.gov (United States)

    So, Jiyeon; Jeong, Se-Hoon; Hwang, Yoori

    2017-04-01

    The extant empirical research examining the effectiveness of statistical and exemplar-based health information is largely inconsistent. Under the premise that the inconsistency may be due to an unacknowledged moderator (O'Keefe, 2002), this study examined a moderating role of outcome-relevant involvement (Johnson & Eagly, 1989) in the effects of statistical and exemplified risk information on risk perception. Consistent with predictions based on elaboration likelihood model (Petty & Cacioppo, 1984), findings from an experiment (N = 237) concerning alcohol consumption risks showed that statistical risk information predicted risk perceptions of individuals with high, rather than low, involvement, while exemplified risk information predicted risk perceptions of those with low, rather than high, involvement. Moreover, statistical risk information contributed to negative attitude toward drinking via increased risk perception only for highly involved individuals, while exemplified risk information influenced the attitude through the same mechanism only for individuals with low involvement. Theoretical and practical implications for health risk communication are discussed.

  12. Risk-informed design guidance for future reactor systems

    International Nuclear Information System (INIS)

    Delaney, Michael J.; Apostolakis, George E.; Driscoll, Michael J.

    2005-01-01

    Future reactor designs face an uncertain regulatory environment. It is anticipated that there will be some level of probabilistic insights in the regulations and supporting regulatory documents for Generation-IV nuclear reactors. Central to current regulations are design basis accidents (DBAs) and the general design criteria (GDC), which were established before probabilistic risk assessments (PRAs) were developed. These regulations implement a structuralist approach to safety through traditional defense in depth and large safety margins. In a rationalist approach to safety, accident frequencies are quantified and protective measures are introduced to make these frequencies acceptably low. Both approaches have advantages and disadvantages and future reactor design and licensing processes will have to implement a hybrid approach. This paper presents an iterative four-step risk-informed methodology to guide the design of future-reactor systems using a gas-cooled fast reactor emergency core cooling system as an example. This methodology helps designers to analyze alternative designs under potential risk-informed regulations and to anticipate design justifications the regulator may require during the licensing process. The analysis demonstrated the importance of common-cause failures and the need for guidance on how to change the quantitative impact of these potential failures on the frequency of accident sequences as the design changes. Deliberation is an important part of the four-step methodology because it supplements the quantitative results by allowing the inclusion in the design choice of elements such as best design practices and ease of online maintenance, which usually cannot be quantified. The case study showed that, in some instances, the structuralist and the rationalist approaches were inconsistent. In particular, GDC 35 treats the double-ended break of the largest pipe in the reactor coolant system with concurrent loss of offsite power and a single

  13. Equipment for inspection of austenitic stainless steel pipe welds

    International Nuclear Information System (INIS)

    Boehmer, W.D.; Horn, J.E.

    1979-01-01

    A computer controlled ultrasonic scanning system and a data acquisition and analysis system have been developed to perform the inservice inspection of welds in stainless steel sodium piping in the Fast Flux Test Facility. The scanning equipment consists of a six axis motion mechanism and control system which allows full articulation of an ultrasonic transducer as it follows the circumferential pipe welds. The data acquisition and analysis system consists of high speed ultrasonic waveform digitizing equipment, dedicated processors to perform on-line analysis, and data storage and display equipment

  14. 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

  15. A new importance measure for risk-informed decision making

    International Nuclear Information System (INIS)

    Borgonovo, E.; Apostolakis, G.E.

    2000-01-01

    Recently, several authors pointed out that the traditional importance measures had limitations. In this study, the problem through an analysis at the parameter level was investigated and a new measure was introduced. The measure was based on small parameter variations and is capable of accounting for the importance of a group of components/parameters. The definition, computational steps, and an application of a new importance measure for risk-informed decision making were presented here. Unlike traditional importance measures, differential importance measure (DIM) deals with changes in the various parameters that determine the unavailability/unreliability of a component, e.g., failure rates, common-cause failure rates, individual human errors. The importance of the component unavailability/unreliability can be calculated from the importance of the parameters. DIM can be calculated for the frequency of initiating events, while risk achievement worth (RAW) is limited to binary events, e.g., component unavailability. The changes in parameters are 'small'. This is more realistic than the drastic assumption in RAW that the component is always down. DIM is additive. This allows the evaluation of the impact of changes, such as the relaxation of quality assurance requirements, which affect groups of parameters, e.g., the failure rates of a group of pumps. (M.N.)

  16. Study on default setting for risk-informed regulation

    International Nuclear Information System (INIS)

    Jang, S.C.; Ha, J.J.; Jung, W.D.; Jeong, K.S.; Han, S.H.

    1998-12-01

    Both performing and validating a detailed risk analysis of a complex system are costly and time-consuming undertakings. With the increased use of probabilistic safety analysis (PSA) in regulatory decision making, both regulated parties and regulators have generally favored the use of defaults, because they can greatly facilitate the process of performing a PSA in the first place as well as the process of reviewing and verifying the PSA. The use of defaults may also ensure more uniform standards of PSA quality. However, regulatory agencies differ in their approaches to the use of default values, and the implications of these differences are not yet well understood. Moreover, large heterogeneity among licensees makes it difficult to set suitable defaults. This study focus on the development of model for setting defaults in order to achieve more applicability of risk-informed regulation. In particular, explored are the effects of different levels of conservatism in setting defaults, and their implications for the crafting of regularity incentives. (author). 17 refs., 1 tab

  17. Background risk information to assist in risk management decision making

    International Nuclear Information System (INIS)

    Hammonds, J.S.; Hoffman, F.O.; White, R.K.; Miller, D.B.

    1992-10-01

    The evaluation of the need for remedial activities at hazardous waste sites requires quantification of risks of adverse health effects to humans and the ecosystem resulting from the presence of chemical and radioactive substances at these sites. The health risks from exposure to these substances are in addition to risks encountered because of the virtually unavoidable exposure to naturally occurring chemicals and radioactive materials that are present in air, water, soil, building materials, and food products. To provide a frame of reference for interpreting risks quantified for hazardous waste sites, it is useful to identify the relative magnitude of risks of both a voluntary and involuntary nature that are ubiquitous throughout east Tennessee. In addition to discussing risks from the ubiquitous presence of background carcinogens in the east Tennessee environment, this report also presents risks resulting from common, everyday activities. Such information should, not be used to discount or trivialize risks from hazardous waste contamination, but rather, to create a sensitivity to general risk issues, thus providing a context for better interpretation of risk information

  18. Risk Informed Optimization of Nuclear Instrumentation for Emergency Conditions

    International Nuclear Information System (INIS)

    Knoll, Alexander

    2013-06-01

    Emergency conditions after a nuclear accident are different in each case and cannot be predicted accurately. The accident at TMI did not contaminate the environment. The accident at Chernobyl had a large, early release of nuclear contamination, widely dispersed over many countries. Although there was no large, early release of contamination at Fukushima Daiichi, the timeline of the accident included days of later contamination of various degrees of severity. A large amount of the contamination has been released to the ocean and an exclusion zone still exists around the station. In all of these accidents there were no adequate radiation monitoring systems distant from the origin point that could provide accurate status to the authorities and the local population. In the recent years a number of new nuclear monitoring systems have been implemented or are under development to be installed in areas that might be exposed to nuclear contamination in emergencies. Based on the risk informed optimization methodology, this paper provides recommendations for selecting the quantity and type of instrumentation, the location and sampling of data, and the real-time processing of information. (authors)

  19. The development of a nuclear accident risk information system

    International Nuclear Information System (INIS)

    Jeong, J. T.; Jeong, W. D.

    2001-01-01

    The computerized system NARIS (Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analyses of the distribution of the health effects. The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system

  20. 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

  1. Development of high-sensitivity ultrasonic techniques for in-service inspection of nuclear reactors

    International Nuclear Information System (INIS)

    Linzer, M.; Parks, S.I.; Dietz, D.R.; Norton, S.J.

    1979-11-01

    During the period covered by this report, the expanding-aperture annular array system and an analysis of the focusing properties of an annular array were completed. Experiments were also undertaken to test the efficacy of spatial-temporal averaging to improve the signal-to-grain noise ratio in austenitic steel

  2. Studies in nondestructive testing with potential for in-service inspection of LMFBRs

    International Nuclear Information System (INIS)

    McClung, R.W.

    1976-01-01

    A variety of nondestructive examination techniques have been and are being developed at ORNL with potential for ISI in LMFBRs. Among these are radiographic techniques for radiation environment and image enhancement, advanced eddy-current techniques and equipment for flaw detection and thickness measurement and ISI of steam generator tubing, and ultrasonic methods for quantitative flaw evaluation using frequency-analysis and bore-side ultrasonic techniques for steam generator tubing. Further developments should result in positive application to ISI

  3. Development of high-sensitivity ultrasonic techniques for in-service inspection of nuclear reactors

    International Nuclear Information System (INIS)

    Linzer, M.

    1977-01-01

    The principal objective of the program is to develop techniques to enhance the sensitivity of ultrasonic signals which are below the random noise of the system. A secondary objective is to develop instrumentation for improved discrimination of flaw signals from background ''clutter'' and for characterization of failure-related material properties through measurements of ultrasonic parameters such as velocity and attenuation. The improved techniques will be applied to detect flaws in nuclear reactor materials and components

  4. Field experience with advanced inservice inspection NDE-techniques for detection and sizing

    International Nuclear Information System (INIS)

    Engl, G.; Kronig, M.

    1988-01-01

    This document deals with Non-Destructive Examination (NDE) techniques used for the detection and sizing of cracks. Several techniques, such as L-SAFT, ALOK and Phased Array with UT-Tomography are discussed and compared. (TEC)

  5. Mobile computer tomography for in-service inspection in nuclear power plants

    International Nuclear Information System (INIS)

    Mueller, J.; Kiechle, H.; Ewert, U.; Raedel, C.; Redmer, B.

    2006-01-01

    COMPRA GmbH, a member of the RTD Group, developed together with BAM Berlin the TomoCAR system for the mechanized digital radiography and planar X-ray tomography. The system is able to accurately size cracks in welds with a COD (crack opening displacement) of only 25 μm independent of its orientation. The system has a very high reliability of recognizing planar defects and the integrated TomoCar software modules are able to produce a 3D-view of the cracks. It is the first time these kinds of results are achieved on site with radiography equipment. (orig.)

  6. Inservice inspection a preventative measure for a utility to improve availability

    International Nuclear Information System (INIS)

    Hausermann, R.

    1985-01-01

    The wish of an everlasting good performance of a machine is as old as the dream to invent the perpetum mobile. The real technical world is different in that the material in use for a structure or a machine depends greatly on the fabrication process, the environment and the loads within the period of use of the equipment. In this paper a method is discussed how, by applying a well balanced maintenance strategy coupled with NDT (Non Destructive Testing), the utility goal, to reach a high work availability, and the authority goal to keep a high safety readiness, can be achieved. The aspect of the NDT is discussed in more detail

  7. A new fracture mechanics approach for defining defect acceptance standards for inservice inspection

    Energy Technology Data Exchange (ETDEWEB)

    Heliot, J [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France); Chevanne, H; Grandemange, J; Barthelet, B [Electricite de France (EDF), 75 - Paris (France). Service de la Production Thermique

    1988-12-31

    This document presents several standards and methods defined by the french national utility Electricite De France (EDF). Methods to predict crack propagation and crack stability in ductile materials are presented together with standards of allowable defects. (TEC).

  8. A new fracture mechanics approach for defining defect acceptance standards for inservice inspection

    International Nuclear Information System (INIS)

    Heliot, J.; Chevanne, H.; Grandemange, J.; Barthelet, B.

    1988-01-01

    This document presents several standards and methods defined by the french national utility Electricite De France (EDF). Methods to predict crack propagation and crack stability in ductile materials are presented together with standards of allowable defects. (TEC)

  9. In-service inspection of sub-coating defects in PWR reactor vessel tubes

    International Nuclear Information System (INIS)

    Birac, A.; Frappier, J.C.; Saglio, Robert.

    1982-08-01

    Since the presence of cracks under the coating of the tubes of certain PWR reactor vessels were noted during manufacture, the need emerged to develop a nondestructive testing method to guarantee the detection of existing cracks and to determine their potential evolution. An ultrasonic testing method was developed for the purpose. In Part 1, the choice of ultrasonic transducers is justified from the theoretical and practical standpoints. In Part 2, the results obtained on test specimens containing artificial defects are presented in accordance with the different parameters involved. In Part 3, covering parts with a large number of real defects, the results of real defect/recorded signal correlations are given, with respect to both detection and dimensions. Examples of automatic data processing are analyzed [fr

  10. In-Service Inspection Approaches for Lead-Cooled Nuclear Reactors

    Science.gov (United States)

    2017-06-01

    heavily regulated and mature. For example, the Illinois Emergency Management Agency (IEMA) conducted 805 soil samples testing for radionuclides around... radiation , and lead-cooled reactors are expected to have economic advantages compared to other nuclear coolant/moderator systems due to design...their six nuclear reactors in 22 2015 (IEMA, 2016, 3). In addition, they currently have 1649 environmental dosimeters testing for gamma radiation

  11. Material degradation analysis and maintenance decisions based on material condition monitoring during in-service inspections

    International Nuclear Information System (INIS)

    Yacout, A.M.; Orechwa, Y.

    1996-03-01

    The degradation of the material in critical components is shown to be an effective measure which can be used to compute the risk adjusted economic penalty associated with different maintenance decisions. The approach of estimating the probability, with confidence interval, of the time that a prescribed degradation level is exceeded is shown to be practical, as demonstrated in the analysis of irradiated fuel cladding. The methodology for the estimation of the probability is predicated on the existence of a parsimonious and robust mixed-effects model of the evolution of the degradation. This model, in general, relates measured surrogates of the degradation level to computed or measured variables, which characterize the environment during the operating history of the component. We propose and demonstrate the efficacy of using an artificial neural network, constructed via a genetic supervisor, as an aid in developing the requisite mixed-effects model and testing its continued validity as new data are obtained

  12. Field experience with advanced inservice inspection NDE-techniques for detection and sizing

    Energy Technology Data Exchange (ETDEWEB)

    Engl, G; Kronig, M

    1988-12-31

    This document deals with Non-Destructive Examination (NDE) techniques used for the detection and sizing of cracks. Several techniques, such as L-SAFT, ALOK and Phased Array with UT-Tomography are discussed and compared. (TEC).

  13. Sampling programs for pre-service and in-service inspection

    International Nuclear Information System (INIS)

    Goodman, J.

    1987-01-01

    The paper outlines methods for practising engineers and statisticians: 1) Classification of sampling procedures and methods; 2) choice of sampling method; 3) criteria of acceptable quality; 4) evaluation of biased samples and nonuniform populations. (HP)

  14. Nitrogen alloyed material for steam turbine and generator components: an alternative to expensive nondestructive inservice inspections

    International Nuclear Information System (INIS)

    Stein, G.; Menzel, J.

    1990-01-01

    In efforts to improve the efficiency of energy conversion systems the trend in power generation is towards increased output capacities, pressures and, in particular, temperatures. If these goals are to be achieved, it will be necessary for materials displaying distinctly improved properties to be used without this involving any dramatic rise in costs, e.g. by the use of superalloys. A new steel making technique, based on pressurized electroslag remelting, has spawned a new generation of advanced engineering materials, the nitrogen alloyed steels. The paper describes the properties of these ferritic and austenitic N-steels and their potential for future applications

  15. 49 CFR 214.529 - In-service failure of primary braking system.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false In-service failure of primary braking system. 214... Maintenance Machines and Hi-Rail Vehicles § 214.529 In-service failure of primary braking system. (a) In the event of a total in-service failure of its primary braking system, an on-track roadway maintenance...

  16. Assessing the In-Service Needs of Basic School Natural Science ...

    African Journals Online (AJOL)

    ... that no significant associations existed. It was recommended among other things that in-service training courses be used as platforms in upgrading the teachers' knowledge and skills. Additionally, in-service course organizers should first assess the in-service needs of participants before providing the appropriate support ...

  17. An Examination of Technology Training Experiences from Teacher Candidacy to In-Service Professional Development

    Science.gov (United States)

    Williams, Mable Evans

    2017-01-01

    The purpose of this qualitative study was to explore the perceptions of in-service teachers concerning the effectiveness of technology training from a teacher education preparation program to in-service professional development. The findings of the study revealed that inservice teachers have had varying degrees of technology experiences from their…

  18. Increasing the Effectiveness of Inservice Training for Desegregation: A Synthesis of Current Research. Reference & Resource Series.

    Science.gov (United States)

    Smylie, Mark A.; Hawley, Willis D.

    This report reviews recent research on strategies that have been found to promote useful and effective inservice training programs in desegregated schools. The first section presents approaches for planning and implementing inservice training for desegregation. The second section describes inservice desegregation training programs that focus on:…

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. Remote inspection of a buried pipeline using a mobile ultrasonic testing system

    Energy Technology Data Exchange (ETDEWEB)

    Muralidhar, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Rajendran, S; Ramakumar, M S [Bhabha Atomic Research Centre, Mumbai (India). Division of Remote Handling and Robotics

    1994-12-31

    The nuclear reactor, Cirus, has now been in operation for three decades. As part of a programme to ascertain the integrity and safety of the various reactor parts in-service inspection of the embedded portion of the main coolant pipeline will be carried out. A mobile ultrasonic testing system has been developed and tested in the laboratory to measure the wall thickness of an underground pipe from the inner corroded surface using a water-bubbler technique. 3 figs.

  3. Inspection during the starting-up phase of nuclear power plants in Spain

    International Nuclear Information System (INIS)

    Figueras Clavijo, J.M.

    1977-01-01

    At present a construction permit in the frame of the Spanish Nuclear Programme has been granted for seven nuclear power plants and three of these have reached an advanced construction stage. The first part of the paper deals with the technical procedures to be followed for commissioning a nuclear power plant according to the provisions of the 1972 regulations on nuclear and radioactive installations. The second part discusses the problems related to conduct of in-service inspections. (NEA) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  5. Automated Inspection of Aircraft

    Science.gov (United States)

    1998-04-01

    This report summarizes the development of a robotic system designed to assist aircraft inspectors by remotely deploying non-destructive inspection (NDI) sensors and acquiring, processing, and storing inspection data. Carnegie Mellon University studie...

  6. Optimising import phytosanitary inspection

    NARCIS (Netherlands)

    Surkov, I.

    2007-01-01

    Keywords: quarantine pest, plant health policy, optimization, import phytosanitary inspection, ‘reduced checks’, optimal allocation of resources, multinomial logistic regression, the Netherlands World trade is a major vector of spread of quarantine plant pests. Border phytosanitary inspection

  7. Approaches to cancer assessment in EPA's Integrated Risk Information System.

    Science.gov (United States)

    Gehlhaus, Martin W; Gift, Jeffrey S; Hogan, Karen A; Kopylev, Leonid; Schlosser, Paul M; Kadry, Abdel-Razak

    2011-07-15

    The U.S. Environmental Protection Agency's (EPA) Integrated Risk Information System (IRIS) Program develops assessments of health effects that may result from chronic exposure to chemicals in the environment. The IRIS database contains more than 540 assessments. When supported by available data, IRIS assessments provide quantitative analyses of carcinogenic effects. Since publication of EPA's 2005 Guidelines for Carcinogen Risk Assessment, IRIS cancer assessments have implemented new approaches recommended in these guidelines and expanded the use of complex scientific methods to perform quantitative dose-response assessments. Two case studies of the application of the mode of action framework from the 2005 Cancer Guidelines are presented in this paper. The first is a case study of 1,2,3-trichloropropane, as an example of a chemical with a mutagenic mode of carcinogenic action thus warranting the application of age-dependent adjustment factors for early-life exposure; the second is a case study of ethylene glycol monobutyl ether, as an example of a chemical with a carcinogenic action consistent with a nonlinear extrapolation approach. The use of physiologically based pharmacokinetic (PBPK) modeling to quantify interindividual variability and account for human parameter uncertainty as part of a quantitative cancer assessment is illustrated using a case study involving probabilistic PBPK modeling for dichloromethane. We also discuss statistical issues in assessing trends and model fit for tumor dose-response data, analysis of the combined risk from multiple types of tumors, and application of life-table methods for using human data to derive cancer risk estimates. These issues reflect the complexity and challenges faced in assessing the carcinogenic risks from exposure to environmental chemicals, and provide a view of the current trends in IRIS carcinogenicity risk assessment. Copyright © 2011. Published by Elsevier Inc.

  8. A new risk-informed design and regulatory process

    International Nuclear Information System (INIS)

    Apostolakis, George E.; Golay, Michael W.; Camp, Allen L.; Duran, Felicia A.; Finnicum, David; Ritterbusch, Stanley E.

    2001-01-01

    The overall purpose of the new approach, termed Risk-Informed Regulation, is to formulate a method of regulation that is logically consistent and devised so that both the reactor designer and regulator can work together in obtaining systems able to produce economical electricity safely. In this new system the traditional tools (deterministic and probabilistic analyses, tests and expert judgement) and treatments (defense-in-depth, conservatism) of safety regulation would still be employed, but the logic governing their use would be reversed from the current treatment. In the new treatment, probabilistic risk analysis (PRA) would be used as the paramount decision support tool, taking advantage of its ability to integrate all of the elements of system performance and to represent the uncertainties in the results. The latter is the most important reason for this choice, as the most difficult part of safety regulation is the treatment of uncertainties, not the assurance of expected performance. The scope of the PRA would be made as large as that of the reactor system, including all of its performance phenomena. The models and data of the PRA would be supported by deterministic analytical results, and data to the extent feasible. However, as in the current regulatory system, the models and data of the PRA would require being complemented by subjective judgements where the former were inadequate. All of these elements play important roles in the current decision-making structure; the main departure from current practice would be making all of these treatments explicit within the PRA, therefore, decreasing the frequency of sometimes arbitrary judgments. In the intended sense the PRA would be used as a vehicle for stating the beliefs of the designer and regulatory decision-maker; the foundation of their decisions. Thus, the PRA should be viewed as a Bayesian decision tool, and be used in order to take advantage of its capabilities in integration and inclusion of

  9. Approaches to cancer assessment in EPA's Integrated Risk Information System

    International Nuclear Information System (INIS)

    Gehlhaus, Martin W.; Gift, Jeffrey S.; Hogan, Karen A.; Kopylev, Leonid; Schlosser, Paul M.; Kadry, Abdel-Razak

    2011-01-01

    The U.S. Environmental Protection Agency's (EPA) Integrated Risk Information System (IRIS) Program develops assessments of health effects that may result from chronic exposure to chemicals in the environment. The IRIS database contains more than 540 assessments. When supported by available data, IRIS assessments provide quantitative analyses of carcinogenic effects. Since publication of EPA's 2005 Guidelines for Carcinogen Risk Assessment, IRIS cancer assessments have implemented new approaches recommended in these guidelines and expanded the use of complex scientific methods to perform quantitative dose-response assessments. Two case studies of the application of the mode of action framework from the 2005 Cancer Guidelines are presented in this paper. The first is a case study of 1,2,3-trichloropropane, as an example of a chemical with a mutagenic mode of carcinogenic action thus warranting the application of age-dependent adjustment factors for early-life exposure; the second is a case study of ethylene glycol monobutyl ether, as an example of a chemical with a carcinogenic action consistent with a nonlinear extrapolation approach. The use of physiologically based pharmacokinetic (PBPK) modeling to quantify interindividual variability and account for human parameter uncertainty as part of a quantitative cancer assessment is illustrated using a case study involving probabilistic PBPK modeling for dichloromethane. We also discuss statistical issues in assessing trends and model fit for tumor dose-response data, analysis of the combined risk from multiple types of tumors, and application of life-table methods for using human data to derive cancer risk estimates. These issues reflect the complexity and challenges faced in assessing the carcinogenic risks from exposure to environmental chemicals, and provide a view of the current trends in IRIS carcinogenicity risk assessment.

  10. Technical objectives of inspection

    International Nuclear Information System (INIS)

    Sorenson, R.J.; Stewart, K.B.; Schneider, R.A.

    1976-01-01

    The various technical objectives of inspection are discussed in a very general manner. The discussion includes how the inspection function is related to the assumed threat, the various degrees of assurance and reliance on criteria, and the hierarchy of assurance which is obtained from the various types or levels of inspection

  11. Software Formal Inspections Guidebook

    Science.gov (United States)

    1993-01-01

    The Software Formal Inspections Guidebook is designed to support the inspection process of software developed by and for NASA. This document provides information on how to implement a recommended and proven method for conducting formal inspections of NASA software. This Guidebook is a companion document to NASA Standard 2202-93, Software Formal Inspections Standard, approved April 1993, which provides the rules, procedures, and specific requirements for conducting software formal inspections. Application of the Formal Inspections Standard is optional to NASA program or project management. In cases where program or project management decide to use the formal inspections method, this Guidebook provides additional information on how to establish and implement the process. The goal of the formal inspections process as documented in the above-mentioned Standard and this Guidebook is to provide a framework and model for an inspection process that will enable the detection and elimination of defects as early as possible in the software life cycle. An ancillary aspect of the formal inspection process incorporates the collection and analysis of inspection data to effect continual improvement in the inspection process and the quality of the software subjected to the process.

  12. GIS In-Service Teacher Training Based on TPACK

    Science.gov (United States)

    Hong, Jung Eun; Stonier, Francis

    2015-01-01

    This article introduces the geographic information systems (GIS) in-service teacher training, focusing on the intersection of technological, pedagogical, and content knowledge (TPACK) for successful implementation of GIS in the classroom. Eleven social studies teachers in Georgia learned GIS technologies, inquiry-based learning, and social studies…

  13. Gender Diversity and ICT Literacy among Nigerian In-Service ...

    African Journals Online (AJOL)

    The instrument used for this study was the self-designed ICT Literacy Scale (ILS) with a coefficient of 0.91. Mean and Standard Deviation, t-test, Pearson Product Moment Correlation Coefficient were utilized for the data analysis. The findings revealed that there gender is a determining factor of in-service teachers' overall ...

  14. 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...

  15. Australian Primary In-Service Teachers' Conceptions of Geography

    Science.gov (United States)

    Preston, Lou

    2015-01-01

    This paper reports on the second part of a two pronged qualitative investigation that examines the ways in which Australian primary teachers conceptualise geography and geography teaching. In the first part of the project, 47 pre-service primary teachers were surveyed. In this paper, I draw on interviews with six in-service primary teachers to…

  16. Structural condition assessment of in-service wood

    Science.gov (United States)

    Robert J. Ross; Brian K. Brashaw; Xiping Wang

    2006-01-01

    Wood is used extensively for both interior and exterior applications in the construction of a variety of structures (residential, agricultural, commercial, government, religious). The deterioration of an in-service wood member may result from a variety of causes during the life of a structure. It is important, therefore, to periodically assess the condition of wood...

  17. In-Service Education: A Blueprint for Action

    Science.gov (United States)

    Keller, Arnold J.

    1978-01-01

    Inservice teacher education is the logical catalyst to any process of school improvement but school board members and school administrators must insure that their forthcoming efforts fulfill three important criteria: (1) Shared decision-making responsibilities with recognized representatives of the teaching staff, preferably at the individual…

  18. Inservice Teacher Training: Experiencing German Culture Down Under

    Science.gov (United States)

    Jansen, Louise; Stracke, Elke

    2005-01-01

    In collaboration with the Australian Capital Territory (ACT) Department of Education and Training, the Australian National University has been offering a professional development program for language teachers (called LIFT, or Language Inservice for Teachers) for more than ten years. As the program is specially tailored to meet teachers' current…

  19. Sweet Rejuvenation: Linking In-Service and Teacher Induction.

    Science.gov (United States)

    McNair, Veda; And Others

    This paper describes a collaborative effort between a local education agency (LEA) and an institution of higher education to link inservice education with induction--teachers teaching teachers. The program, based on the Joyce coaching paradigm and recent cognitive development research, posits that long-term training conducted by trained teachers…

  20. Appraisal of in-service trainee teachers' competency on applications ...

    African Journals Online (AJOL)

    Most procedures used to obtain information about student learning are subsumed under assessment. This study appraised in-service trainee teachers' competency on applications of principles of test items writing. One Study Centre was purposively drawn from the National Teachers' Institute for the study. A sample of 33 ...