Repair of Concrete Nuclear Power Plant

Unclassified

NEA/CSNI/R(2002)7/VOL 1

Organisation de Coopération et de Développement Economique sOrganisation for Economic Co-operation and Development

05-Sep-2002

English - Or. Englis hNUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S

OECD-NEA WORKSHOP ON THE EVALUATION OF DEFECTS ,REPAIR CRITERIA & METHODS OF REPAIR FOR CONCRETESTRUCTURES ON NUCLEAR POWER PLANT S

Hosted by GRS at the DIN Institute in Berlin, German y

10th-11th April, 2002

JT00130882

Document complet disponible sur OLIS dans son format d’origineComplete document available on OLIS in its original format


ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30thSeptember 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed :

- to achieve the highest sustainable economic growth and employment and a rising standard of living in Membe rcountries, while maintaining financial stability, and thus to contribute to the development of the world economy ;

- to contribute to sound economic expansion in Member as well as non-member countries in the process of economi cdevelopment ; and

- to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance wit hinternational obligations .

The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece ,Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdo mand the United States . The following countries became Members subsequently through accession at the dates indicated hereafter :Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18t hMay 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996), Korea (12t hDecember 1996) and the Slovak Republic (14th December 2000) . The Commission of the European Communities takes part in th ework of the OECD (Article 13 of the OECD Convention) .

NUCLEAR ENERGY AGENC Y

The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEE CEuropean Nuclear Energy Agency . It received its present designation on 20th April 1972, when Japan became its firs tnon-European full Member. NEA membership today consists of 27 OECD Member countries : Australia, Austria, Belgium ,Canada, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg ,Mexico, the Netherlands, Norway, Portugal, Republic of Korea, Spain, Sweden, Switzerland, Turkey, the United Kingdom and th eUnited States . The Commission of the European Communities also takes part in the work of the Agency .

The mission of the NEA is:

- to assist its Member countries in maintaining and further developing, through international co-operation, th escientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclea renergy for peaceful purposes, as well as

- to provide authoritative assessments and to forge common understandings on key issues, as input to governmen tdecisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainabl edevelopment .

Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive wast emanagement, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law an dliability, and public information . The NEA Data Bank provides nuclear data and computer program services for participatingcountries.

In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency inVienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field .

©OECD 200 2Permission to reproduce a portion of this work for non-commercial purposes or classroom use should be obtained through th eCentre français d’exploitation du droit de copie (CCF), 20, rue des Grands-Augustins, 75006 Paris, France, Tel. (33-1) 44 07 4770, Fax (33-1) 46 34 67 19, for every country except the United States . In the United States permission should be obtained throug hthe Copyright Clearance Center, Customer Service, (508)750-8400, 222 Rosewood Drive, Danvers, MA 01923, USA, or CC COnline : http://www.copyright.com/. All other applications for permission to reproduce or translate all or part of this book shoul dbe made to OECD Publications, 2, rue André-Pascal, 75775 Paris Cedex 16, France .

2


COMMITTEE ON NUCLEAR REGULATORY ACTIVITIE S

The Committee on Nuclear Regulatory Activities (CNRA) of the OECD Nuclear Energy Agency (NEA) i san international committee made up primarily of senior nuclear regulators . It was set up in 1989 as a forum for theexchange of information and experience among regulatory organisations and for the review of developments whichcould affect regulatory requirements.

The Committee is responsible for the programme of the NEA, concerning the regulation, licensing an dinspection of nuclear installations . The Committee reviews developments which could affect regulatory requirement swith the objective of providing members with an understanding of the motivation for new regulatory requirementsunder consideration and an opportunity to offer suggestions that might improve them or avoid disparities amongMember Countries . In particular, the Committee reviews current practices and operating experience .

The Committee focuses primarily on power reactors and other nuclear installations currently being builtand operated . It also may consider the regulatory implications of new designs of power reactors and other types ofnuclear installations .

In implementing its programme, CNRA establishes co-operative mechanisms with NEA's Committee o nthe Safety of Nuclear Installations (CSNI), responsible for co-ordinating the activities of the Agency concerning th etechnical aspects of design, construction and operation of nuclear installations insofar as they affect the safety of suc hinstallations . It also co-operates with NEA's Committee on Radiation Protection and Public Health (CRPPH) an dNEA's Radioactive Waste Management Committee (RWMC) on matters of common interest.

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S

The NEA Committee on the Safety of Nuclear Installations (CSNI) is an international committee made upof scientists and engineers . It was set up in 1973 to develop and co-ordinate the activities of the Nuclear EnergyAgency concerning the technical aspects of the design, construction and operation of nuclear installations insofar a sthey affect the safety of such installations . The Committee's purpose is to foster international co-operation in nuclea rsafety amongst the OECD Member countries .

CSNI constitutes a forum for the exchange of technical information and for collaboration betwee norganisations which can contribute, from their respective backgrounds in research, development, engineering o rregulation, to these activities and to the definition of its programme of work . It also reviews the state of knowledgeon selected topics of nuclear safety technology and safety assessment, including operating experience . It initiates andconducts programmes identified by these reviews and assessments in order to overcome discrepancies, developimprovements and reach international consensus in different projects and International Standard Problems, and assist sin the feedback of the results to participating organisations . Full use is also made of traditional methods of co-operation, such as information exchanges, establishment of working groups and organisation of conferences andspecialist meeting.

The greater part of CSNI's current programme of work is concerned with safety technology of wate rreactors . The principal areas covered are operating experience and the human factor, reactor coolant systembehaviour, various aspects of reactor component integrity, the phenomenology of radioactive releases in reacto raccidents and their confinement, containment performance, risk assessment and severe accidents . The Committeealso studies the safety of the fuel cycle, conducts periodic surveys of reactor safety research programmes and operate san international mechanism for exchanging reports on nuclear power plant incidents .

In implementing its programme, CSNI establishes co-operative mechanisms with NEA's Committee o nNuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation ,licensing and inspection of nuclear installations with regard to safety . It also co-operates with NEA's Committee o nRadiation Protection and Public Health and NEA's Radioactive Waste Management Committee on matters o fcommon interest.

3

NEA/CSNI/R(2002)7/VOL1

4


Foreword

The Committee on the Safety of Nuclear Installations (CSNI) of the OECD-NEA co-ordinates the NE Aactivities concerning the technical aspects of design, construction and operation of nuclear installation sinsofar as they affect the safety of such installations . In 1994, the CSNI approved a proposal to set up aTask Group under its Principal Working Group 3 (recently re-named as the Working Group on Integrity o fComponents and Structures (IAGE)) to study the need for a programme of international activities in th earea of concrete structural integrity and ageing and how such a programme could be organised . The taskgroup reviewed national and international activities in the area of ageing of nuclear power plant concretestructures and the relevant activities of other international agencies . A proposal for a CSNI programme ofworkshops was developed to address specific technical issues which were prioritised by OECD-NEA taskgroup into three levels of priority :

First Priority

• Loss of prestressing force in tendons of post-tensioned concrete structure s• In-service inspection techniques for reinforced concrete structures having thick sections and area s

not directly accessible for inspection

Second Priority

• Viability of development of a performance based database• Response of degraded structures (including finite element analysis techniques )

Third Priority

• Instrumentation and monitoring• Repair methods• Criteria for condition assessment

The working group has progressively worked through the priority list developed during the preliminar ystudy carried out by the Task Group . Currently almost all of the three levels of priority are effectivelycomplete, although in doing so the committee has identified other specific items worthy of consideration .By working logically through the list of priorities the committee has maintained a clarity of purpose whic hhas been important in maintaining efficiency and achieving its objectives . The performance of the grouphas been enhanced by the involvement of regulators, operators and technical specialists in both the work o fthe committee and its technical workshops and by liaison and co-operation with complementar ycommittees of other international organisations . The workshop format that has been adopted (based aroun dpresentation of pre-prepared papers or reports followed by open discussion and round-table development o frecommendations) has proved to be an efficient mechanism for the identification of best practice, potentia lshortcomings of current methods and identification of future requirements .

5


6


SUMMARY

OECD-NEA workshop on the evaluation of defects, repair criteria & methods of repair for concret estructures on nuclear power plant s

OECD-NEA IAGE held an international workshop on the evaluation of defects, repair criteria & method sof repair for concrete structures on nuclear power plants in Berlin, Germany on April 10-11, 2002 .Through 2 technical sessions devoted to Operational Experience and State of the Art and Futur eDevelopments, a broad picture of the status was given to a large audience composed by 54 participant sfrom 17 countries and International Organisations . 21 papers have been presented at the Workshop .

The objectives of the workshop were to examine the current practices and the state of the art with regard t othe evaluation of defects, repair criteria and methods of repair for concrete structures on Nuclear Powe rPlants with a view to determining the best practices and identification of shortfalls in the current methods,which are presented in the form of conclusions and recommendations in this report .

This workshop on the evaluation of defects, repair criteria and methods of repair for concrete structures o nNuclear Power Plants is the latest in a series of workshops .

The complete list of CSNI reports, and the text of reports from 1993 on, is available onhttp://www.nea .fr/html/nsd/docs/

7


8


Acknowledgement

Gratitude is expressed to GRS, Germany for hosting the Workshop at the DIN Institute in Berlin. Inparticular, special thanks to Mr . Helmut Schulz and Dr Jurgen Sievers, and also Mrs Brunhilde Laue andMrs Schneider for their help .

Thanks are also expressed to chairmen of the sessions and to the Organizing Committee for their effort an dco-operation.

Dr Leslie M Smith BEG(UK) Ltd (UK) ChairmanProf Pierre Labbé IAEA (International )M. Jean-Pierre Touret EdF (France)Herr Rüdiger Danisch Framatome ANP GmbH (Germany)Mr James Costello USNRC (USA)Dr Dan Naus ORNL (USA)M.Eric Mathet OECD-NEA (International)

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10


OECD-NEA WORKSHOPON THE

EVALUATION OF DEFECTS, REPAIR CRITERIA & METHODS OF REPAIR FO RCONCRETE STRUCTURES ON NUCLEAR POWER PLANTS

10th and 11 th April, 2002Berlin, Germany

A. CONTENTSB. CONCLUSIONS AND RECOMMENDATIONSC. PROGRAMMED. PAPER SE. PARTICIPANTS

11


12


A. TABLE OF CONTENTS

PAGE

Volume 1

B. CONCLUSIONS AND RECOMMENDATIONSC. PROGRAMMED. PAPER S

Introductory Paper

Inspection, Assessment and Repair of Nuclear Power Plan tConcrete StructuresD. J . Naus, Oak Ridge National Laboratory, U.S.A .H. L. Graves, J.F. Costello, USNRC, U.S.A .

Repair of the Gentilly-1 Concrete Containment Structur eA. Popovic, D . Panesar and M. Elgohary, AECL, Canada

The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures an dInstallation of an Automated Cathodic Protection Syste mL. M. Smith, C . A. Hughes, British Energy Generation UK, Ltd .G. Jones, Sea-Probe, Ltd .

Feasibility Study of IE-SASW Method for the Non-Destructive Evaluation of Containmen tBuilding of Nuclear Power PlantMr. Yong-Pyo Suh, KEPRI, Korea

Field Studies of Effectiveness of Concrete Repair sN.J.R. Baldwin, Mott MacDonald Ltd.,(UK)

Detection and Repair of Defects in the Confinement Structures at Paks NP PMr. Nyaradi Csaba, Paks NPP Ltd

Steam Generator Replacement at Ringhals 3 Containment, Transport Openin gJan Gustavsson, Ringhals Nuclear Power Plant, (Sweden )

In Service Inspection Programme and Long Time Monitoring of Temelin NP PContainment StructuresJan Maly, Jan Stepan, Energoprojekt Prague, Czech Republi c

Repair Criteria and Methods of Repair for Concrete Structures on Nuclear Power Plants

135R. Lasudry, Tractebel Energy Engineering, (Belgium )

SESSION A: OPERATIONAL EXPERIENCEChairman: Mr. Rüdiger Danisch, Framatome-ANP GmbH (Germany) 37

SESSION A: OPERATIONAL EXPERIENCE (Continued )Chairman: Dr James Costello, USNRC (USA)

115

171923

25

39

5 1

59

73

83

117

125

13


Post-Fire Damage Assessment Procedures for Nuclear Power Plant Structures

157L.M. Smith, British Energy Generation UK, Ltd ., (UK)

Volume 2

SESSION B: STATE OF THE ART & FUTURE DEVELOPMENTS

17Chairman: Dr Naus, ORNL (US)

Various stages to Address Concrete Cracking on NPP sC. Seni, Mattec Engineering Ltd ., (Canada)

Investigation of the Leakage Behaviour of Reinforced Concrete Wall sNico Herrmann, Christoph Niklasch, Michael Stegemann, Lothar Stempniewski ,University of Karlsruhe, (Germany )

The Development of a State-of-the-Art Structural Monitoring Instrumentation Syste mfor Nuclear Power Plant Concrete StructuresL. M. Smith, B . Stafford, M.W. Roberts, British Energy Generation UK ,Ltd.A. McGown, University of Strathclyde (UK)

Ageing and Static Reliability of Concrete Structures under Temperatur eand Force LoadingPetr Stepanek,, Stanislav Stastnik Vlastislav Salajka, Technical University of Brno ,Jaroslav Skolai, Jiri Stastny, Dukovany Power Plant, (Czech Republic )

Efficient Management of Inspection and Monitoring Data for a Bette rMaintenance of InfrastructureMarcel de Wit, Gilles Hovhanessian, Advitam

Aging Process Of A Good Concrete During Forty Year sDr. Peter Lenkei, University, College of Engineering (Hungary)

SESSION B: State of the Art & Future Developments (Continued)

8 1Chairman: Mr. Jean-Pierre TOURET, EdF, (France)

Acoustic Monitorin gMarcel de Wit, Gilles Hovhanessian, Advitam

Concrete Properties Influenced by Radiation Dose During Reactor Operatio nTakaaki Konno, Secretariat of Nuclear Safety Commission, (Japan )

The Use of Composite Materials in the Prevention and Strengthening of Nuclea rConcrete StructuresD. Chauvel, P .A. Naze, J-P . Touret EdF, Villeurbanne, (France)

19

3 1

4 1

55

67

77

83

97

105

14


Detection of Reinforcement Corrosion and its Use for Service Life Assessment o fConcrete StructuresC. Andrade, I . Martinez, J . Munoz, CSIC (SP) Rodr iguez, M. Ramirez, GEOCISA (Spain)

Improved Detection of Tendon Ducts and Defects in Concrete Structures Usin gUltrasonic Imagin gW. Müller, V. Schmitz, FIZP, (GE) M. Krause, M . Wiggenhauser, Bundesanstalt fürMaterialforschung und -Prüfung, (Germany )

Structural Integrity Evaluation of a Steel Containment for the Replacement o fSteam GeneratorMr. Yong-Pyo Suh, KEPRI (Korea)

New Methods on Reconstruction of Safety Compartments of Nuclear Power Plants

141Z. Kdpper, Kdpper und Partner, Bochum, (Germany)D. Busch, RWE Solutions AG, Essen, (Germany)

E.

PARTICIPANTS

151

115

125

133

15


16


OECD-NEA Workshop on the Evaluation of Defects, Repair Criteria &Methods of Repair for Concrete Structures on Nuclear Power Plants ,

GRS, Berlin, Germany April 10-11, 200 2

Conclusions and Recommendations

The objectives of the workshop were to examine the current practices and the state of the art with regard t othe evaluation of defects, repair criteria and methods of repair for concrete structures on Nuclear Powe rPlants with a view to determining the best practices and identification of shortfalls in the current methods,which are presented in the form of conclusions and recommendations .

CONCLUSIONS

1. Repairs to concrete NPP structures and their durability will continue to be an issue until finaldecommissioning .

2. Experience gained during repair projects, and extensive field studies of repaired structures, show tha tthe effectiveness of the concrete repair is dependent on :

correct diagnosis of the cause of the damage ;selection of a repair strategy that addresses this cause ;choice of appropriate repair materials and methods ;careful management of the process ;post repair maintenance strategy supported by comprehensive records .

Computerised databases can assist with : recording the detection and diagnosis of damage ; rexcording thelocation of, and specification for, repairs ; and management of the subsequent repair .

3. The combination of concrete with composite materials is useful in a repair situation . These materialsnow have a track record in structural repairs to a decommissioned prestressed concrete containment(PCC). Extensive testing has proved their potential as an alternative to steel as a liner for PCCs . Theyare currently being considered for enhancing the leak tightness of unlined containments .

4. Surface overcoating materials can protect exposed concrete surfaces from deterioration due toenvironmental factors eg carbonation, chlorides etc . Careful design will ensure that the coating systemcan accommodate structural movement, maximise durability and satisfy aesthetic considerations .

5. Experiences of repairs, supported by field studies of repaired structures, confirm that a principal caus eof damage to reinforced concrete structures is corrosion of the reinforcement . Impressed currentcathodic protection (CP) has been shown to be effective in improving the durability of a repaire dstructure exposed to a very severe marine environment . Laboratory examination of samples of concret eremoved from a structure protected by CP has shown that long term application of impressed currentCP was not detrimental to the original concrete and did not affect the steel to concrete bond .

6. There was recognition that the nuclear industry might benefit from improved guidance on assessmen tof defects and the effectiveness of subsequent repairs . However, absolute criteria are difficult to defin eand may not be universally applicable .

7. Laboratory trials of impact echo and synthetic aperture focusing technique ultrasonic non-destructiv etesting have shown that they have potential to detect subsurface features in concrete elements but thatsignificant further development is required for field implementation .

17


8. Laboratory tests are providing important data on the leakage of air and air/steam through cracke dconcrete . These tests will help to inform the assessment of pressure retaining concrete structures o nNPPs and provide a useful source of validation for numerical models and simulation .

9. Papers on the application of structural monitoring to NPP concrete structures confirmed that th epracticality of installing instrumentation is of equal importance to its ability to measure the damag eparameter under investigation . Acoustic monitoring has demonstrated the potential to detect and locatecracking and may warrant further consideration as a tool for assisting the testing of containmentstructures .

10. There is little data available on the effect of irradiation on concrete . Samples of concrete removed froma biological shield structure have provided some information .

11. A pre-prepared structural condition assessment procedure listing nuclear safety related structures ma ybe useful in assessing post-fire damage on NPPs . Materials used in the repair of fire damage must becapable of meeting the fire performance criteria required by the original structure .

RECOMMENDATION S

The following recommendations are offered to inform national activities and research programmes for theinspection, maintenance and repair of concrete NPP structures .

1. The execution and durability of repairs to concrete should be considered as an issue that is relevant t othe nuclear safety of NPPs throughout the lifetime of the plant and until final decommissioning .

2. Improved guidance is required on the assessment of defects (eg . cracks) and the performance andeffectiveness of subsequent repairs .

3. Owners/operators of NPP concrete structures should develop procedures for recording: the detectionand diagnosis of defects/damage ; the location of and specification for each repair ; and the managementstrategy to be applied to the repair.

4. Further development of NDE techniques is required in order to support the assessment and evaluationof defects and subsequent repairs in concrete structures . The development priorities, conclusions andrecommendations identified at the OECD-NEA 1997 Risley NDE workshop should be applied .

5. Further work is required on the evaluation of leakage through cracks in concrete structures .

6. Further investigation of the effects of irradiation on concrete is required .

7. Consideration should be given to the development of pre-prepared structural condition assessmen tprocedures listing nuclear safety related structures for the evaluation of post-fire damage on NPPs .

18



EVALUATION OF DEFECTS, REPAIR CRITERIA & METHODS OF REPAIR FORCONCRETE STRUCTURES ON NUCLEAR POWER PLANTS

Hosted by GRS at the DIN Institute in Berlin, German y10 & 11 APRIL 200 2

C. PROGRAMME

Wednesday April 10 , 2002

9:30 10:00 Welcome

H. Schulz, GRSE. Mathet - OECD

Introduction

L. Smith- Chairman

Introductory paper

10:00 10:30 Inspection, Assessment and Repair of Nuclear Power Plant D. J. Naus, Oak Ridge NationalConcrete Structures

Laboratory, U.S.A .H. L. Graves, J . F . Costello,USNRC, U.S.A .

10:30 10:50 Coffee break

SESSION A: Operational ExperienceChairman :

Mr. Ruediger.Danisch, FRAMATOME-ANP GmbH (GE )

10:50 11:10 Repair of the Gentilly-1 Concrete Containment Structure

A. Popovic, D . Panesar and M .Elgohary, AECL, (CDN )

11 :10 11:30 The Repair of Nuclear Power Plant Reinforced ConcreteMarine Structures and Installation of an Automated Cathodi cProtection System

L. M . Smith, C.A. Hughes, BritishEnergy Generation UK ,Ltd .G. Jones, Sea-Probe, Ltd .

19


11 :30 11:50 Feasibility Study of IE-SASW Method for the Non-Destructive Evaluation of Containment Building of NuclearPower Plant

Mr. Yong-Pyo Suh, KEPRI (K)11 :50 12:10 Field Studies of Effectiveness of Concrete Repairs

N.J.R. Baldwin, Mott MacDonaldLtd.,(UK)

12:10 12:45 Detection and repair of defects in the confinement structure sat Paks NPP

Mr. Nyaradi Csaba, Paks NPPLtd

12:45 14:00 Lunch

SESSION A: Operational Experience (Continued)Chairman: Dr James Costello, USNRC (USA)

14:00 14:20

Steam Generator Replacement at Ringhals 3 Containment ,Transport Opening

Jan Gustavsson, Ringhals NuclearPower Plant, (SW)

14:20 14:40 In

Service

Inspection

Programme

and

Long

Time

Jan Maly, Jan Stepan ,Energoprojekt Prague, CzechRepublic

Monitoring of Temelin NPP Containment Structure s

14:40 15:00 Repair Criteria and Methods of Repair for ConcreteStructures on Nuclear Power Plants

R. Lasudry, Tractebel EnergyEngineering, (BE )

15:00 15:20 Post-Fire Damage Assessment Procedures for Nuclea rPower Plant Structures

L.M. Smith, British EnergyGeneration UK ,Ltd., (UK)


20


SESSION B: State of the Art & Future Developments

15:50Chairman :

16:10 Various stages to address Concrete Cracking on NPPsDr Naus, ORNL (US )

16:10

C. Seni, Mattec Engineering Ltd. ,(CDN)

16:30

Investigation of the Leakage Behaviour of

Reinforced

16:30

Concrete WallsNico Herrmann, ChristophNiklasch, Michael Stegemann ,Lothar Stempniewski,University of Karlsruhe, (GE)

16:50

The

Development

of

a

State-of-the-Art

Structural

16:50 17:10

Monitoring Instrumentation System for Nuclear Power Plan tConcrete Structures

Ageing and Static Reliability of Concrete Structures under

L. M. Smith, B . Stafford, M .W.Roberts, British EnergyGeneration UK ,Ltd.A. McGown, University ofStrathclyde (UK)

17:10 17:30

Temperature and Force LoadingPaper not presented but in the proceedings

Efficient management of inspection and monitoring data for

Petr Stepanek,, StanislavStastnik Vlastislav Salajka ,Technical University of Brno ,Jaroslav Skolai, Jiri Stastny ,Dukovany Power Plant, (CZ )

17:30 17:45

a better maintenance of infrastructure

Aging process of a good concrete during forty years

Marcel de Wit, GillesHovhanessian, Advitam

End of the first day

Dr. Peter Lenkei, University,College of Engineering(Hungary)

Thursday April 11, 200 2


Chairman :

9:00 9:30 Acoustic monitoring

Mr. Jean-Pierre TOURET, EdF, (F)

Marcel de Wit, GillesHovhanessian, Advitam

2 1


9:30 9:50 Concrete Properties Influenced by Radiation Dose DuringReactor Operation

Takaaki Konno, Secretariat ofNuclear Safety Commission, (J)

9:50 10:10 The Use of Composite Materials in the Prevention andStrengthening Of Nuclear Concrete Structures

D. Chauvel, P .A. Naze, J-P .Touret EdF, Villeurbanne, (F)

10:10 10:30 Detection of reinforcement corrosion and its use for servicelife assessment of concrete structures

C. Andrade, I . Martinez, J .Munoz, CSIC (SP) Rodriguez ,M. Ramirez, GEOCISA (SP )


11 :00 11:20 Improved Detection of Tendon Ducts and Defects inConcrete Structures Using Ultrasonic Imaging

W. Müller, V. Schmitz, FIZP,(GE) M. Krause, M .Wiggenhauser, Bundesanstal tfür Materialforschung und -Prüfung, (GE)

11 :20 11:40 Structural Integrity Evaluation of a Steel Containment forthe Replacement of Steam Generator

Mr. Yong-Pyo Suh, KEPRI(KR)

11 :40 12:00 New Methods on Reconstruction of Safety Compartments o fNuclear Power Plants

Z. Kdpper, Kdpper und Partner,Bochum, (GE)D. Busch, RWE Solutions AG,Essen, (GE)

12:00 14:00 Lunch

SESSION C Chairman:

Dr L.M. Smith (UK)14:00 16:30 Panel discussion

16:30

Closure

22


C. PAPERS

23


24


INSPECTION, ASSESSMENT, AND REPAIR O FNUCLEAR POWER PLANT CONCRETE STRUCTURE S

D.J. NausOak Ridge National Laboratory (ORNL)

Oak Ridge, TN

H.L. Graves, III and J .F. CostelloU.S. Nuclear Regulatory Commission (USNRC)

Washington, D .C .

Abstract

Aging of concrete structures occurs with the passage of time and has the potential, if its effects ar enot controlled, to increase the risk to public health and safety . Activities have been conducted to addres sfactors related to quantifying the effects of age-related degradation on nuclear power plant concret estructures and components . Environmental effects that can lead to age-related degradation of reinforcedconcrete structures and their manifestations are described. Current regulatory in-service testing andinspection requirements are reviewed. Techniques commonly used to inspect nuclear power plant concret estructures to assess and quantify age-related degradation are identified. An approach for conduct o fcondition assessments is presented, as well as criteria, based on visual indications, for use in classificationand assessment of concrete degradation. Materials and techniques for repair of reinforced concretestructures are noted and guidance provided on repair options available for various forms of concret edegradation (e .g., cracking, spalling, and steel reinforcement corrosion) . Nuclear power plant degradationand repair experience is summarized. In-service inspection/repair strategies to maintain the probability o ffailure of a concrete component at or below a target value are discussed .

1 .

Introduction

To date, 104 nuclear power reactors are currently licensed for commercial operation in the Unite dStates (US) . Currently 103 of these reactors are in operation, producing about 20% of the nation’ selectricity supply . The median age of these reactors is over 20 years, with 61 having been in commercia loperation for 20 or more years . Initial operating licenses for these reactors will start expiring in 2006, wit happroximately 10% expiring by the year 2010, and more than 40% by the year 2015 . Continuing theservice of existing nuclear power plants (NPPs) through a renewal of their initial operating license sprovides a timely and cost-effective solution to the problem of meeting future electricity demand. In fact,48 reactor units (as of March 2002) have either completed the license renewal process, submittedapplications to renew their operating licenses, or announced that they intend to do so . However, thestructures in these plants are susceptible to aging by various processes, depending on the operatin genvironment and service conditions, that can affect the engineering properties, structuralresistance/capacity, failure mode, and location of failure initiation . As a result, the ability of the structure sto withstand various challenges in service from operating conditions, the natural environment, andaccidents may be impacted . Current aging-related activities in large measure are therefore focusingtechnical development and support on condition assessment with the aim of demonstrating that structuralmargins of existing plants have not or will not erode during the desired service life due to aging orenvironmental effects . Probabilistic methods can be used to provide the quantitative tools for theassessment of uncertainty in condition assessment and are an essential ingredient of risk-informedmanagement decisions concerning continued service of the NPP structures .

25


2. NPP Concrete Containments

All commercial NPPs contain concrete structures whose performance and function are necessary fo rprotection of the safety of plant operating personnel and the general public . Most of the concrete structuresin NPPs are similar to conventional civil engineering structures; however, certain NPP concrete structure scan entail thicker sections, increased reinforcement, and limited accessibility and harsher exposur econditions at certain locations . Typical safety-related functions that the concrete structures provide includefoundation, support, shielding, and containment. Although a number of concrete structures are importan tto the overall safety of NPPs (e .g ., fuel/storage pools, cooling water intake structures, and foundations) ,discussions will concentrate on the containment structures because of their unique requirements .

Concrete containments are metal lined, reinforced concrete pressure-retaining structures that in som ecases may be post-tensioned. The concrete vessel includes the concrete shell and shell components, shel lmetallic liners, and penetration liners that extend the containment liner through the surrounding shellconcrete . The reinforced concrete shell, which generally consists of a cylindrical wall with a hemisphericalor ellipsoidal dome and flat base slab, provides the necessary structural support and resistance to pressure-induced forces . Leak-tightness is provided by a steel liner fabricated from relatively thin plate material(e.g., 6-mm thick) that is anchored to the concrete shell by studs, structural steel shapes, or other stee lproducts . Initially, existing building codes, such as American Concrete Institute (ACI) Standard 318 ,Building Code Requirements for Reinforced Concrete [1], were used in the nuclear industry as the basi sfor design and construction of concrete structural members . However, because the existing building code sdid not cover the entire spectrum of design requirements and because they were not always considere dadequate, additional criteria were developed for design of seismic Category 1 (i .e ., safety related)structures (e .g ., definitions of load combinations for both operating and accident conditions) . Plants thatused early ACI codes for design were reviewed by the USNRC through the Systematic Evaluation Progra mto determine if there were any unresolved safety concerns [2] . Current rules for construction of concretecontainments are provided in Section III, Division 2 of the ASME Code [3] . The USNRC has developedsupplemental load combination criteria and provides information related to concrete and steel interna lstructures of steel and concrete containments [4,5] . Rules for design and construction of the metal linerthat forms the pressure boundary for the reinforced concrete containments are found in Section III ,Division 1, Subsection NE of the ASME Code . Depending on the functional design (e .g., large dry or ic econdenser), NPP concrete containments can be on the order of 40 to 50 m diameter and 60 to 70 m high ,with wall and dome thicknesses from 0 .9 to 1 .4 m, and base slab thicknesses from 2 .7 to 4 .1 m. Almostthree-quarters of the NPPs licensed for commercial operation in the US employ either a reinforced concret eor post-tensioned concrete containment . Boiling-water reactor plants in the US that utilize a stee lcontainment have reinforced concrete structures that serve as secondary containments or reactor building sthat provide support and shielding functions for the primary containment .

3. Potential Degradation Factors

Degradation is considered to be any phenomenon that decreases the load-carrying capacity of thecontainment, limits its ability to contain a fluid medium, or reduces its service life . Service-relateddegradation can affect the ability of a NPP containment to perform satisfactorily in the unlikely event of asevere accident . The root cause for component degradation can generally be linked to a design o rconstruction problem, inappropriate material application, a base- or weld-metal flaw, maintenance orinspection activities, or a severe service condition . Primary mechanisms or factors that can producepremature deterioration of concrete structures include those that impact either the concrete or reinforcin gsteel materials (i .e ., mild steel reinforcement or post-tensioning system) . Degradation of concrete can b ecaused by adverse performance of its cement-paste matrix or aggregate materials under either chemical o rphysical attack. Chemical attack may occur in several forms : efflorescence or leaching, sulfate attack

26


(including delayed ettringite formation), attack by acids and bases, salt crystallization, and alkali-aggregat ereactions . Physical attack mechanisms for concrete include freeze/thaw cycling, thermal exposure/therma lcycling, abrasion/erosion/cavitation, irradiation, and fatigue or vibration . Degradation of mild steelreinforcing materials can occur as a result of corrosion, irradiation, elevated temperature, or fatigue effects .Post-tensioning systems are susceptible to the same degradation mechanisms as mild steel reinforcement ,plus loss of prestressing force, primarily due to tendon relaxation and concrete creep and shrinkage .

4.

Testing and Inspection Requirement s

One of the conditions of all operating licenses for water-cooled power reactors is that the primaryreactor containment shall meet the requirements set forth in Appendix J, "Primary Reactor Containmen tLeakage Testing for Water-Cooled Power Reactors," to 10 CFR Part 50 [6] . Contained in Appendix J arerequirements pertaining to Type A, B, and C leakage-rate tests that must be performed by each licensee a sa condition of their operating license. On September 26, 1995, the USNRC amended Appendix J (60 FR49495) to provide a performance-based option for leakage-rate testing as an alternative to the existingprescriptive requirements . The amendment is aimed at improving the focus of the body of regulations byeliminating prescriptive requirements that are marginal to safety and by providing licensees greaterflexibility for cost-effective implementation methods for regulatory safety objectives .

Appendix J to 10 CFR Part 50 requires a general inspection of the accessible interior and exterio rsurfaces of the containment structures and components to uncover any evidence of structural deterioratio nthat may affect either the containment structural integrity or leak-tightness . On August 8, 1996, theUSNRC published an amendment (61 FR 41303) to 10 CFR 50 .55a of its regulations to require thatlicensees use portions of the ASME Code for containment in-service inspection . Specifically, the rulerequires that licensees adopt the 1992 Edition with the 1992 Addenda of Subsection IWE, "Requirement sfor Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants," andSubsection IWL, "Requirements for Class CC Concrete Components of Light-Water Cooled Powe rPlants," of Section XI . In addition, several supplemental requirements with respect to the concrete andmetal containments were included in the rule . On September 22, 1999 the USNRC again amended 10 CF RPart 50.55a to endorse use of the 1995 Edition up to and including 1996 Addenda of Section XI ,Subsections IWE and IWL, of the ASME Code for inspection of containment structures . Subsequently onAugust 3, 2001, the USNRC announced that it intends to amend 10 CFR Part 50.55a to incorporate byreference the 1997 Addenda, the 1998 Edition, the 1999 Addenda, and the 2000 Addenda of Section XI o fthe ASME Code [7] . Comments on the proposed amendment are presently being addressed .

5.

In-Service Inspection and Condition Assessment

Operating experience has demonstrated that periodic inspection, maintenance, and repair ar eessential elements of an overall program to maintain an acceptable level of reliability over the service lif eof a NPP containment, or in fact, of any structural system . Knowledge gained from conduct of an in-service condition assessment can serve as a baseline for evaluating the safety significance of an ydegradation that may be present, and defining subsequent in-service inspection programs and maintenanc estrategies .

Effective in-service condition assessment of a containment requires knowledge of the expected typ eof degradation, where it can be expected to occur, and application of appropriate methods for detecting an dcharacterizing the degradation. Degradation detection is the first and most important step in the conditio nassessment process . Routine observation, general visual inspections, leakage-rate tests, and nondestructiv eexaminations are techniques used to identify areas of the containment that have experienced degradation .Techniques for establishing time-dependent change such as section thinning due to corrosion, or changes i ncomponent geometry and material properties, involve monitoring or periodic examination and testing.

27


Knowing where to inspect and what type of degradation to anticipate often requires information about th edesign features of the containment as well as the materials of construction and environmental factors .Basic components of the continued service evaluation process for NPP concrete structures include damag edetection and classification, root-cause determination, and measurement .

5 .1

In-service inspection

Nondestructive test methods are used to determine hardened concrete properties and to evaluate thecondition of concrete in structures . 1 Application of these methods for detection of degradation i nreinforced concrete structures involves either a direct or indirect approach . The direct approach generallyinvolves a visual inspection of the structure, removal/testing/analysis of material, or a combination of th eabove. Indirect approaches measure some property of concrete (e .g., rebound number or ultrasonic puls evelocity) and relate it to strength, elastic behavior, or extent of degradation through correlations that hav ebeen established previously. Many of the nondestructive test methods are based on the indirect approach,in which a small number of destructive and nondestructive tests are conducted in tandem at noncritica llocations in a structure to develop the required correlation curve(s) . However, destructive tests may not b epossible in many areas of a NPP structure to develop the required curves so assessment of in-place strengt hmust be based on published relations . Environment-specific methods are used where surfaces of structure sare not accessible for direct inspection due to the presence of soils, protective coatings, or portions o fadjacent structures . These methods provide an indirect assessment of the physical condition of th estructure (i.e ., potential for degradation) by quantifying the aggressiveness of the environment adjacent tothe structure (e .g ., air, soil, and groundwater) . If results of these tests indicate that the environmentadjacent to the structure is not aggressive, one might conclude that the structure is not deteriorating .However, when conditions indicate that the environment is potentially conducive to degradation, additiona lassessments are required that may include exposure of the structure for visual or limited destructive testing .

5 .2 Condition assessment

Determining the existing performance characteristics and extent and causes of any observed distres sis accomplished through a condition assessment . Common in the condition assessment approaches is th econduct of a field survey, involving visual examination and application of nondestructive and destructivetesting techniques, followed by laboratory and office studies . Guidelines and direction on conduct o fsurveys of existing general civil engineering buildings are available [10,11] . The condition survey usuallybegins with a review of the "as-built" drawings and other information pertaining to the original design an dconstruction so that information, such as accessibility and the position and orientation of embedded stee lreinforcing and plates in the concrete, is known prior to the site visit. Next is a detailed visual examinationof the structure to document easily obtained information on instances that can result from or lead t ostructural distress. Visual inspections are one of the most valuable of the condition survey method sbecause many of the manifestations of concrete deterioration appear as visible indications o rdiscontinuities on exposed concrete surfaces . Visual inspections encompass a variety of techniques (e .g . ,direct and indirect inspection of exposed surfaces, crack and discontinuity mapping, physicaldimensioning, environmental surveying, and protective coatings review) . To be most effective, the visualinspection should include all exposed surfaces of the structure ; joints and joint materials ; interfacingstructures and materials (e .g ., abutting soil) ; embedments ; and attached components (e.g ., base plates andanchor bolts) . Degraded areas of significance are measured . The condition of the surrounding structure sshould also be examined to detect occurrence of differential settlement or note aggressiveness of the loca loperating environment . Results obtained should be documented and photographs or video images taken of

1 Descriptions and principles of operation, as well as applications, for nondestructive test methods most commonl yused to determine material properties of hardened concrete in existing construction and to determin estructural properties and assess conditions of concrete are available [8,9] .

28


any discontinuities and pertinent findings . A crack survey is usually done by drawing the locations an dwidths of cracks on copies of project plans . Cracking patterns may appear that suggest weaknesses in theoriginal design, construction deficiencies, unanticipated thermal movements, chemical reactivity ,detrimental environmental exposure, restrained drying shrinkage, or overloading . Distress associated withcracks such as efflorescence, rust stains, or spalling is noted . After the visual survey has been completed ,the need for additional surveys such as delamination plane, corrosion, or pachometer is determined. Thedelamination plane survey is used to identify internal delaminations that are usually caused by corrosion o fembedded metals or internal vapor pressure. Results of the visual and delamination surveys are used toselect portions of the structure that will be studied in greater detail. To locate areas of corrosion activitywithin reinforced concrete, copper-copper sulfate half-cell studies can be performed . By taking readings atmultiple locations on the concrete surface, an evaluation of the probability of corrosion activity o fembedded reinforcing steel (or other metals) can be made . Where significant chloride penetration i ssuspected, concrete powder samples or cores should be removed from several depths extending to an dbeyond the embedded outer layer of reinforcing steel . Also, a pachometer survey may be performed a spart of the detailed study to confirm the location of steel reinforcement. Where there is evidence of severecorrosion, the steel bar should be uncovered to allow visual inspection and measurement of cross-sectiona larea loss . Upon return to the office, results of the field survey are evaluated in detail . A crack survey mapis prepared and studied for meaningful patterns . Half-cell data are studied and isopotential lines are drawnto assist in determining active corrosion sites . Samples of concrete and steel obtained from area sexhibiting distress are tested in the laboratory . Chloride ion results are plotted versus depth to determinethe profile and the chloride content at the level of the steel . Any elements that appear to be structurall ymarginal, due either to unconservative design or deterioration effects, are identified and appropriat ecalculation checks made . These analyses may identify distress in the structure that has been caused bystructural overload and indicate safety factors . If the calculations are inconclusive, suitable load testin gmay be indicated (if feasible) . After all of the field and laboratory results have been collated and studiedand all calculations have been completed, a report is prepared .

Cracking is a very common damage by-product from a large number of concrete degradatio nmechanisms. Active concrete cracking is difficult to assess in terms of impact on structural behavior and isdifficult to repair . Thus, inspection methods that support the early identification, sizing, and cause o fcracking in concrete structures are of primary interest for future inspections . Also, the primary concern forall metallic constituents of concrete structures is corrosion and corrosion-related damage . Inspections thatidentify early signs of corrosion cell initiation and indicate the rate of propagation are similarly valuable .A visual-based approach based primarily on the results of visual inspections has been developed fo rassistance in the classification and treatment of conditions or findings that might emanate from in-servic einspections of reinforced concrete structures . '

The visual-based approach uses a "three-tiered" hierarchy” so that through use of different levels o facceptance, minor discontinuities can be accepted and more significant degradation in the form of defect scan be evaluated in more detail [13,14] . The three acceptance levels include acceptance without furthe revaluation, acceptance after review, and additional evaluation required. Criteria associated with theseacceptance levels are presented elsewhere [13] . Evaluations under these acceptance levels can involveextensive application of both nondestructive and destructive testing methods and detailed analytica levaluations frequently may be required to better characterize the current condition of the structure an dprovide the basis for formulation of a repair strategy (if needed) . Even if the analysis results indicate tha tthe component is acceptable at present, additional assessments should be conducted to demonstrate that th ecomponent will continue to meet its functional and performance requirements during the desired servic e

Information is also available on a damage-based approach that is founded on the concept that degradation of acomponent in service is manifested in physical evidence (i .e ., measurable values) that can be categorized orclassified into distinct stages or conditions in accordance with their impact on performance [12] .

29


life (i .e ., take into account the current structural condition and use service life models to estimate the futur eimpact of pertinent degradation factors on performance) .

6 .

Repair of Reinforced Concrete

Reinforced concrete structures can start to deteriorate due to exposure to the environment (e .g . ,temperature, moisture, and cyclic loading) almost from the time of construction [15]. The rate ofdeterioration is dependent on the component’s structural design, materials selection, quality of construction,curing, and aggressiveness of its environmental exposure . Termination of a component’s service life occur swhen it can no longer meet its functional and structural requirements . Results provided through periodi capplication of in-service inspection techniques as part of a condition assessment program can be used todevelop and implement a remedial action prior to the structure achieving an unacceptable level o fperformance. Depending on the degree of deterioration and the residual strength of the structure, th efunction of a remedial measures activity may be structural, protective, cosmetic, or any combination ofthese .

6.1 Repair considerations

The first step in any repair activity is a thorough assessment of the damaged structure or componen tincluding evaluation of the (1) cause of deterioration, (2) extent of deterioration, and (3) effect ofdeterioration on the functional and performance requirements of the structure or component . From thisinformation a remedial measures strategy is developed based on the consequence of damage (e.g., affect ofdegradation on structural safety), time requirements for implementation (e .g ., shutdown requirements ,immediate or future safety concern), economic aspects (e .g., partial or complete repair), and residua lservice life requirements (e.g., desired residual service life will influence action taken) [16] . Basicremedial measures options include (1) no active intervention ; (2) more frequent inspections or conduct ofspecific studies; (3) if safety margins are presently acceptable, take action to prevent deterioration fro mgetting worse; (4) carry out repairs to restore deteriorated or damaged parts of structure to a satisfactorycondition ; and (5) demolish and rebuild all or part of structure . Quite often options (3) and (4) ar econsidered jointly .

6.2 Repair materials and techniques

Deterioration of reinforced concrete generally will result in cracking, spalling, or delamination o fthe cover concrete . Corrosion resulting from either carbonation or the presence of chlorides is th edominant type of distress that impacts reinforced concrete structures . More detailed information to thatprovided below on typical remedial measures for NPP concrete structures is available [16-18] .

After identifying that a crack is of sufficient size to require repair, it is important to determine i fthe crack is dormant or active (i .e ., mechanism still operating) . Dormant cracks can be resin injected usingepoxy or high molecular weight methacrylate (HMWM) . Active cracks must be treated as if they ar econtrol joints and require special treatment, especially if fluid leakage is involved . Surface preparation iscritical to a successful spall repair . The concrete substrate must be sound and the exposed surface dry an dfree of oil, grease, and loose particles . The most appropriate materials for patching are those that ar eclosest in composition to the material to be patched . Usually this means portland cement concrete for larg epatches or portland cement mortar for small ones ; however, non-portland cement binders have been usedsuccessfully. By patching with a cementitious material, the final thermal and structural properties of th erepair will be similar to the base concrete. Where the repairs are exposed to aggressive fluids the chemicalcomposition of the fluids should be known and the repair materials must be compatible . Delamination scan be repaired by removal and replacement of the delaminated concrete . In areas where removal o fconcrete is not required, the delaminated area can be repaired by injection of epoxy or HMWM . Proper

30


surface preparation, batching, mixing, placing, and curing are all important for long-term durability o fconcrete repairs. Basic repair solutions for corrosion-damaged reinforced concrete include :(1) realkalization by either direct replacement of contaminated concrete with new concrete, use of acementitious material overlay, or application of electrochemical means to accelerate diffusion of alkalisinto carbonated concrete ; (2) limiting the corrosion rate by changing the environment (e .g., drying) toreduce the electrolytic conductivity; (3) steel reinforcement coating (e .g ., epoxy) ; (4) chloride extraction bypassing an electric current (DC) from an anode attached to the concrete surface through the concrete to th ereinforcement (chloride ions migrate to anode) ; and (5) cathodic protection .

6.3 NPP repair experience

A survey was distributed to solicit information on the locations and types of concrete distres scommonly found in US NPPs [17] . Twenty-nine plants representing forty-two reactor units responded .The results of this survey are summarized below :

Concrete structure evaluations are usually limited to an assessment of prestressing systems of post-tensioned concrete containments and a general visual survey of exposed concrete surfaces ;

Twenty-six of the plants reported concrete damage or deterioration with cracking and spallingbeing most common ;

Most common locations of deterioration in BWR plants were in the containment dome and in thewalls and slabs of auxiliary structures, and in PWR plants the locations were in slabs, walls an dequipment supports of reactor buildings and auxiliary structures ;

Twenty-seven of the plants have repaired damaged concrete with epoxy injection, grout injection ,and flexible sealing of cracks being the most common methods utilized ; and

Follow-up evaluation of concrete repairs were not commonly performed .

In general, many of the reported degradation instances associated with the NPP concrete structure soccurred early in the life of the structures and have been attributed to construction/design deficiencies ,improper material selection, or environmental effects . Examples of some of the specific problems thathave occurred due to age-related degradation include concrete containment liner corrosion, leaching o ftendon gallery concrete, corrosion of steel reinforcement in water-intake structures, failure of prestressingtendon wires due to corrosion, and freeze-thaw damage to containment dome . Although the vast majorityof the structures will continue to meet their functional and performance requirements during their servic eperiod, it is reasonable to assume that there will be isolated incidents where the structures may not exhibitthe desired durability without some form of intervention .

7 .

Time-Dependent Reliability

Evaluation of structures for continued service should provide quantitative evidence that thei rcapacity is sufficient to withstand future demands within the proposed service period with a level ofreliability sufficient for public safety . Structural aging will cause the integrity of structures to evolve ove rtime (e .g., a hostile service environment may cause structural strength and stiffness to degrade) .Uncertainties that complicate the evaluation of aging effects arise from a number of sources : inherent

This survey was conducted prior to the ammendment to 10 CFR 50.55a requiring licensees to use Subsections IWEand IWL of the ASME Code for containment in-service inspections, and conduct of inspections of selecte dplants by the USNRC [19] .

31


randomness in structural loads, initial strength, and degradation mechanisms ; lack of in-service inspectionmeasurements and records ; limitations in available databases and models for quantifying time-dependen tmaterial changes and their contribution to structural capacity ; inadequacies in non-destructive evaluation ;and shortcomings in existing methods to account for repair. Any evaluation of the reliability of areinforced concrete structure during its service life must take into account these effects, plus any previou schallenges to the integrity that may have occurred .

Time-dependent reliability analysis methods provide a framework for performing conditio nassessments of existing structures and for determining whether in-service inspection and maintenance ar erequired to maintain reliability and performance at the desired level . The duration of structural loads thatarise from rare operating or environmental events, such as accidental impact, earthquakes, and tornadoes ,is short and such events occupy a negligible fraction of a structure’s service life . Such loads can bemodeled as a sequence of short-duration load pulses occurring randomly in time . The occurrence in timeof such loads is described by a Poisson process, with the mean (stationary) rate of occurrence, 1 randomintensity, Sj , and duration, ,r The number of events, N(t), to occur during service life, t, is described by theprobability mass function,

P (1)

The intensity of each load is a random variable, described by the cumulative distribution functio n(CDF) Fi(x) . In general, the load process is intermittent and the duration of each load pulse has a nexponential distribution,

F Td= 1 - exp[-t/ ti] ; t > 0

(2)

in which i = average duration of the load pulse . The probability that the load process is nonzero at anyarbitrary time is p = 1ti . Loads due to normal facility operation or climatic variations may be modeled b ycontinuous load processes . A Poisson process with rate 1 may be used to model changes in load intensityif the loads are relatively constant for extended periods of time .

The strength, R, of a structural component is described b y

R = B Rm(X 1 , X2, …, Xm)

(3)

in which X 1 , X2 . . . are basic random variables that describe yield strength of steel, compressive or tensil estrength of concrete, and structural component dimensions or section properties . The function Rm(…)describes the strength based on principles of structural mechanics . Modeling assumptions invariably mustbe made in deriving Rm(…) and the factor B describes errors introduced by modeling and scaling effects .The probability distribution of B describes bias and uncertainty that are not explained by the model R m(…)when values of all variables Xi are known. The probability distribution of B can be assumed to be normal .A more accurate behavioral model leads to a decrease in the mean and variability in B and thus in R .Probabilistic models for R usually must be determined from the statistics of the basic variables, X i, since itseldom is feasible to test a sufficient sample of structural components to determine the cumulativ edistribution function (CDF) of R directly .

32


The failure probability of a structural component can be evaluated as a function of (or an interval of )time if the stochastic processes defining the residual strength and the probabilistic characteristics of th eloads at any time are known. The strength, R(t), of the structure and applied loads, S(t), are both rando mfunctions of time . Assuming that degradation is independent of load history, at any time t the margin o fsafety, M(t), is

M(t) = R(t) - S(t) .

(4)

Making the customary assumption that R and S are statistically independent random variables, th e(instantaneous) probability of failure is ,

Pf t ) = P M t ) < 0 (5 )

in which FR(x) and fS(x) are the CDF of R and probability density function (PDF) of S . Equation (5 )provides an instantaneous quantitative measure of structural reliability, provided that Pf(t) can be estimatedand/or validated [20] . It does not convey information on how future performance can be inferred from pastperformance.

For service life prediction and reliability assessment, one is more interested in the probability o fsatisfactory performance over some period of time, say (0,t), than in the snapshot of the reliability of th estructure at a particular time provided by Eq . (5) . Indeed, it is difficult to use reliability analysis fo rengineering decision analysis without having some time period in mind (e .g., an in-service maintenanceinterval) . The probability that a structure survives during interval of time (0,t) is defined by a reliabilityfunction, L(0,t) . If, for example, n discrete loads S 1 , S 2 , . . ., Sn occur at times t 1 , t2, . . ., tn during (0,t), thereliability function becomes ,

L(t) = P[R(t 1 ) > S 1 , … , R(tn) > Sn]

(6)

in which R(ti) = strength at time of loading S i .

Taking into account the randomness in the number of loads and the times at which they occur as wel las initial strength, the reliability function becomes [21 ]

L (t )= f exp4- tl 1 - t -1 f ôFs(g ir~dt fRl 0(r ) dt (7)

in which fR0 = PDF of the initial strength R0 and gi = fraction of initial strength remaining at time of load

S i . The probability of failure during (0,t) i s

F(t) = 1 - L(t) .

(8)

The conditional probability of failure within time interval (t, t+ At), given that the component has survive dup to t, is defined by the hazard function which can be expressed a s

h(t) = -d [ln L(t)]/dt .

(9)

The reliability and hazard functions are integrally relate d

33


L(t) = exp -f ô h (x )dx .

(10)

The hazard function is especially useful in analyzing structural failures due to aging or deterioration . Forexample, if the structure has survived during the interval (0, t 1 ), it may be of interest in scheduling in-service inspections to determine the probability that it will fail before t 2 . Such an assessment can b eperformed if h(t) is known . If the time-to-failure is Tf, this probability can be expressed as

= 1 - exp (f t2 h(xdx .

(11 )P

In turn, the structural reliability for a succession of inspection periods i s

L (0, t)_

L(ti-1, 4- )exp 1~ -ft

t

h (x )dxit

(12)

in which ti-1 = 0 when i = 1 .

Intervals of inspection and maintenance that may be required as a condition for continued operatio ncan be determined from the time-dependent reliability analysis . Forecasts of reliability enable the analys tto determine the time period beyond which the desired reliability of the structure cannot be assured . Atsuch a time, the structure should be inspected. The density function of strength, based on prior knowledg eof the materials in the structure, construction, and standard methods of analysis, is indicated by fR (r) . Theinformation gained during scheduled inspection, maintenance and repair causes the characteristics ofstrength to change ; this is denoted by the (conditional) density fR(r B), in which B is an event dependent onin-service inspection . Information gained from the inspection usually involves several structural variable sincluding dimensions, defects, and perhaps an indirect measure of strength or stiffness . If these variablescan be related through event B, then the updated density of R following in-service inspection is ,

fR r B ) = P r < R :9 r + dr, B P B ]= c K r ) fR r

(13 )

in which fR(r) is termed the prior density of strength, K(r) is denoted the likelihood function, and c is anormalizing constant . The time-dependent reliability analysis then is re-initialized following in-servic einspection/repair using the updated f R(r B) in place of fR (r) . The updating causes the hazard function to b ediscontinuous .

Optimal intervals of inspection and repair for maintaining a desired level of reliability can b edetermined based on minimum life cycle expected cost considerations . Preliminary investigations of suchpolicies have found that they are sensitive to relative costs of inspection, maintenance, and failure [22] . Ifthe cost of failure is an order (or more) of magnitude larger than inspection and maintenance costs, th eoptimal policy is to inspect at nearly uniform intervals of time . However, additional research is requiredbefore such policies can be finalized as part of an aging management plan . Applications of the time-dependent reliability methodology to concrete components are available [22-24] .

8. Conclusions

The performance of reinforced concrete structures in NPPs has been good, reflecting the initial qualit ycontrol, the young age, and the generally benign environment within the a plant . However, as thesestructures age incidences of degradation are likely to increase and if not controlled, degradation has th epotential to reduce the margins that the structures have to withstand various challenges in service from

34


operating conditions, the natural environment, and accidents . The most common form of degradationobserved in NPPs has been concrete cracking. When properly used and applied, in-service inspectiontechniques are effective in detecting aging effects and providing vital input for assessing the condition o fstructures. Methods for conduct of condition assessments of reinforced concrete structures are fairly wel lestablished and generally start with a visual examination of the structure's accessible surfaces . Someguidance has been developed to aid in interpreting results of the condition assessment, but more definitiv ecriteria are required to assist in interpreting the data provided . Repair methods for general civil engineeringreinforced concrete structures are fairly well established and effective when properly implemented, however ,the long-term effectiveness (or durability) of remedial measures require development. Time-dependentreliability analysis methods provide a framework for performing condition assessments of existing structuresand determining whether in-service inspection and maintenance are required to maintain reliability an dperformance at the desired level, however, quantitative data for input into the methodology are limited andthe reliability models for condition assessment have not been validated .

9. Acknowledgements

Research sponsored by the Office of Nuclear Regulatory Research, U .S . Nuclear RegulatoryCommission under Interagency Agreement 1886-N604-3J with the U .S. Department of Energy underContract DE-AC05- 96OR22725 . The submitted paper has been authored in part by a contractor of th eU.S. Government under Contract No . DE-AC05-96OR22725 . This paper has also been prepared in part byan employee of the USNRC and presents information that does not currently represent an agreed-upo nStaff position. The USNRC has neither approved nor disapproved its technical content . The U.S .Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of thi scontribution, or allow others to do so, for U .S . Government purposes .

10. References

1. American Concrete Institute, 1964 . Building Code Requirements for Reinforced Concrete, ACIStandard 318-71, Detroit, Michigan .

2. Lo, T. et al ., 1984. "Containment Integrity of SEP Plants Under Combined Loads," in Proceedings ofthe ASCE Conference on Structural Engineering in Nuclear Facilities, J . Ucciferro (ed.), AmericanSociety of Civil Engineers, New York, New York .

3. American Society of Mechanical Engineers, 2001 . "Rules for Construction of Nuclear Power Plan tComponents," ASME Boiler and Pressure Vessel Code, Sect . III, New York, New York .

4. U.S. Nuclear Regulatory Commission, 1981 . "Concrete Containment," Sect. 3 .8.1 in RegulatoryStandard Review Plan, NUREG-0800, Directorate of Licensing, Washington, DC .

5. U.S. Nuclear Regulatory Commission, 1981 . "Concrete and Steel Internal Structures of Steel andConcrete Containments," Sect . 3 .8 .3 in Regulatory Standard Review Plan, NUREG-0800, Directorateof Licensing, Washington, DC .

6. Office of the Federal Register, 1995 . "Primary Reactor Containment Leakage Testing for Water -Cooled Power Reactors," Appendix J in Code of Federal Regulations, 10 CFR Part 50, Office ofFederal Register, Washington, DC .

7. Office of theFederal Register, 2001 . "Proposed Rules," Vol . 66, No. 150, pp. 40626-40640, Office ofFederal Register, Washington, DC .

8. ACI, 1995. "In-Place Methods for Determination of Strength of Concrete," ACI 228 .1R, AmericanConcrete Institute, Farmington Hills, Michigan .

9. ACI, 1998. "Nondestructive Test Methods for Evaluation of Concrete Structures," ACI 228 .2R,American Concrete Institute, Farmington Hills, Michigan .

10. American Society of Civil Engineers, 1991 . Guidelines for Structural Condition Assessment ofExisting Buildings, ANSI/ASCE 11-90, New York, New York .

11. Perenchio, W. F., 1989. "The Condition Survey," Concrete International, 11(1), pp . 59-62, American

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Concrete Institute, Detroit, Michigan .12. Naus, D. J ., Braverman, J .I., Miller, C .A., Ellingwood, B .R., Hofmayer, C.H., 2000. “Factors Related

to Degradation of Nuclear Power Plant Concrete Structures,” Proc. of International RILEM Workshopon Aging Management and Life Prediction of Concrete Structures, Cannes, France .

13. Hookham, C.J., 1995. “In-Service Inspection Guidelines for Concrete Structures in Nuclear Powe rPlants,” ORNL/NRC/LTR-95/14, Oak Ridge National Laboratory, Oak Ridge, Tennesee .

14. ACI, 1996 . Evaluation of Existing Nuclear Safety-Related Concrete Structures, ACI 349.3R-96,American Concrete Institute, Farmington Hills, Michigan .

15. Browne, R.D., 1989. "Durability of Reinforced Concrete Structures," New Zealand ConcreteConstruction, Parts 1 and 2 .

16. Price, W.F. et al ., 1993 . Review of European Repair Practice for Corrosion Damaged ReinforcedConcrete, Report No. 1303/91/5823, Taywood Engineering Ltd ., R & D Division, London, England .

17. Krauss, P .D., 1994 . Repair Materials and Techniques for Concrete Structures in Nuclear Powe rPlants, ORNL/NRC/LTR-93/28, Martin Marietta Energy Systems, Inc ., Oak Ridge NationalLaboratory, Oak Ridge, Tennessee .

18. Emmons, P .H., 1993 . Concrete Repair and Maintenance Illustrated, R. S. Means Company, Inc . ,Kingston, Massachusetts .

19. Ashar, H. and Bagchi, G., 1995. “Assessment of Inservice Conditions of Safety-Related Nuclea rPower Plant Structures,” NUREG-1522, U.S. Nuclear Regulatory Commission, Washington, DC .

20. Ellingwood, B .R., 1992. “Probabilistic Risk Assessment,” Eng. Safety, McGraw-Hill Book Co ., Ltd. ,London, England, pp. 89-116 .

21. Ellingwood, B.R. and Mori, Y., 1993. “Probabilistic Methods for Condition Assessment and LifePrediction of Concrete Structures in Nuclear Power Plants,” Nuc. Eng. and Des ., Elsevier ScienceS.A., North-Holland, Amsterdam, The Netherlands, Vol 142, pp . 155-166.

22. Mori, Y. and Ellingwood, B.R., 1994. “Maintaining Reliability of Concrete Structures II : OptimumInspection/Repair Strategies,” Ibid, pp . 846-862 .

23. Mori, Y. and Ellingwood, B .R., 1994. “Maintaining Reliability of Concrete Structures I : Role ofInspec-tion/Repair,” J. of Struct. Eng., American Society of Civil Engineers New York, New York ,120(3), pp . 824-845 .

24. Ellingwood, B .R. and Song, J., 1996. “Impact of Structural Aging in Probabilistic Risk Assessment o fReinforced Concrete Structures in Nuclear Power Plants,” NUREG/CR-6425, U.S. Nuclear RegulatoryCommission, Washington, DC .

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SESSION A: OPERATIONAL EXPERIENCEChairman: Mr. Rüdiger Danisch, Framatome-ANP GmbH (Germany )

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38


REPAIR OF THE GENTILLY- 1CONCRETE CONTAINMENT STRUCTURE

A Popovic, D Panesar and M . ElgoharyAtomic Energy of Canada Limite d

Mississauga, Ontario, Canada

Abstract

Gentilly-1 CANDU Nuclear Power Plant is located in Gentilly, Quebec, south of the St . Lawrence River.Gentilly-1 was designed in the late 1960'ies and constructed in the mid 1970'ies . The reactor building ringbeam has suffered concrete degradation for more than fifteen years and has been repaired in 1985 and i n1993. These repairs were ineffective since extensive deterioration had continued to occur . The ring beamcontains and protects the prestressing anchorages and the horizontal ring beam tendons, in addition to th epre-stressing anchorages for the dome prestressing system . In August 1998, an assessment program for theconcrete containment building was initiated . The investigation showed that the structural concrete formingthe dome and the perimeter wall is expected to endure without ageing problems for at least fifty years .However, portions of the ring beam concrete `secondary concrete' needed repair . Various design optionsfor the ring beam repair were considered. Atomic Energy of Canada Ltd . (AECL) and Intelligent Sensingfor Innovating Structures (ISIS) Canada developed a design for the remedial work and a subsequentmonitoring system. The innovative Ring Beam Repair Program was implemented and successfull yperformed in 2000/2001 . The purpose of this paper is to describe the design and field implementation ofthe repair program .

Introduction

The Plant was designed by AECL in the late 1960'ies and became operational in 1972 . In 1980, AECLdecided to place the station in a lay-up state and to generate a plan for its final disposition . It wasconcluded in 1984 that returning the site to a condition of completely unrestricted access and usage was no timmediately necessary and, for technical and financial considerations, the attainment of this objectiv eshould be delayed for the next fifty to eighty years . AECL, the owner of the facility, was required t omaintain the facility in a "static state" . In general, the purpose of the maintaining of the facility was toprovide interim storage for all conventional and radiological hazards until the facility is finall ydecommissioned and demolished . Therefore, it was decided that Gentilly-1 containment structure wouldbe required to function for more than fifty years beyond its originally expected design life . However, thestructure was visibly aging in the ring beam secondary concrete area that was protecting the prestres sanchorages . The poor appearance of the structure was significantly influencing perception of safety an dproper functioning of the nuclear installations . The ring beam deterioration is shown in Figures 1a and 1b .

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Figure 1a, b: G-1 Ring Beam, South Side

Figure 2: Gentilly-1 Containment Building – Installation of Formwork (April 2000 )

A detailed study was undertaken to perform a condition assessment of the structure and the action snecessary to ensure satisfactory performance of the structure over the remaining period of required servic elife . Structural condition assessment studies have been completed and it has been concluded that, excep tthe part of the secondary concrete in the ring beam, the concrete containment structure has the potential t obe serviceable possibly for fifty years or more, as required by AECL .Based on the findings from the condition assessment, recommendations were developed for the repai rdesign and implementation performed on the ring beam to ensure the satisfactory performance of all part sof the concrete containment structure .

40


Condition Assessment and Design Requirement s

As a result of the condition assessment completed in the year 2000, and all previous investigations ,valuable information was gathered based on : visual inspections, in-situ measurements and laborator ytesting and analysis . A comprehensive analysis which included testing of concrete, reinforcing steel andprestressing steel materials, chloride ion content, water soluble alkali content, air voids, water absorption ,compressive strength, modulus of elasticity, Poisson’s ratio, etc ., concluded that the containment structur ewas in good structural condition. It was also concluded that the reactor building could be expected t oremain serviceable for the next fifty years if the ring beam repair is performed .

The repair of the ring beam of the Gentilly-1 containment structure was developed by AECL and ISI SCanada. The objective of the concrete repair was to : to remove all unsound and unbonded concrete ; torestore the concrete of the ring beam ; to protect pre-stressing anchorages; and to improve aesthetics of thebuilding and increase public confidence . The Project had challenges relating to the technical component o fthe repair, as well as to the logistic construction challenges due to the work at the top of the reacto rbuilding, as shown in Figure 2 .

The technical part of the concrete repair project had three main tasks :1. Concrete demolition (removal of unsound concrete) and surface preparation .2. Concrete repair .3. Fiber Reinforced Polymers (FRP) protection of the repaired area .

Concrete demolition, surface preparation and concrete repair are part of the conventional engineering task sand adequate experience and expertise can yield a high standard of repair design and installation .On the other hand, protection of the repaired concrete using FRP is a newer area in the nuclear industr yand does not have an extensive track record . Therefore, a team of specialists was assembled to study an dpropose the technical solution for the application of FRP .

FRP composites have been used for nearly thirty years in aerospace and manufacturing applications wher elow weight, high tensile strength and non-corrosive structural properties are required . In civil engineering ,applications of different types of FRPs are finding their role in fabric roof structures, internal concret ereinforcement, deck grating and as externally bonded reinforcement or protection . The FRP system hasproven benefits in some applications . The technique, known as a wet lay-up, provides flexibility ,constructability and short installation times, resulting in lower overall cost . The system may use differenttype of fibers : carbon, E-glass and Aramid, depending on the particular requirements of an application .The fiber fabric is installed using epoxy resin formulated for substrate adhesion, durability andconstructability.

41


Figure 3 details the application concept of FRP sheets .

Figure 3: FRP - Installation Detai l

42


Figure 4: Ring beam FRP Installation Pattern

PROJECT EXECUTION

Concrete Demolitio n

As shown in Figure 5, using the sounding method, all of the unsound and unbonded concrete was marked,measured and prepared for demolition .

Figure 5: Sounding the Concrete Surface

The unsound concrete was removed using saw cuts and jack hammers, as presented in Figure 6 . Forshallow demolished sections (less than 50 mm deep), the saw cut around the perimeter of repair was at

43


least 12 mm. For deep demolished sections (more than 50 mm deep), the saw cut was at least 25 mm . Themaximum demolition depth into the ring beam was 0 .66 m.

Figure 6: Concrete Demolitio n

After the demolition, and prior to the grouting/concreting, skin reinforcement was installed, Figure 7 .Surface preparation by hydro-jetting or sand blasting was employed to open the pore structure of th econcrete surface and to remove dirt and other debris material, Figure 8 .

Figure 7: Installation of Skin Reinforcement

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Figure 8: Surface Preparation –Hydro-Jetting

Concrete Repair - Materials and Application Method s

The concrete repair materials and repair techniques differ for shallow repairs and deep repairs . Theshallow repair patches have an average thickness of less than 50 mm . The deep repair patches have anaverage thickness of greater than 50 mm .

The shallow patches are repaired with Sika Repair 225 . The repair material is a prepackaged ready to use ,cementitious, high strength, shrinkage compensated mortar, which includes silica fume and fibrereinforcement. The mix is used with the amount of water specified in the technical data sheet . Thematerial was applied by hand troweling .

The deep patches are repaired with Sika Grout 212, SikaCem 810, PeaGravel (5-9 mm) and water . Thepatches are formed, the concrete mix is poured and vibrated while being placed in the forms as shown i nFigures 9 and 10 .

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Figure 9: Deep Repair - Concrete Pour

Figure 10 : Deep Repair – Complete d

46


Glass Fiber Reinforcing Polymer (GFRP) Installatio n

For this application, the MBrace (Master Builder Technologies) Composite Strengthening System wa schosen. Its function is primarily to protect and ensure durability of the concrete repairs . The structuralstrength and high modulus of elasticity were not critical requirements . Based on the material properties, E-glass fiber (EG-900) fabric was the selected material since it exhibited the best flexibility and elasticity o fthe system for this application .

The steps for the GFRP application are as follows :

Step 1 : MBrace Primer (low viscosity to penetrate the concrete pore structure) .Step 2 : MBrace Putty (high viscosity epoxy paste used for surface leveling) .Step 3 : The 1st Resin Coating, MBrace Saturant (low sag epoxy that encapsulates the fibers) .Step 4 : MBrace GFRP EG 900 E-Glass Fiber Fabric (instead of C-Fiber shown in Figure 3) .Step 5 : The 2nd Resin Coating, MBrace Saturant (low sag epoxy that encapsulates the fibers) .Step 6 : Protective Coating: Sonocoat Topcoat Super Colorcoat VOC Top Coat was used .

Site application of the GFRP to the ring beam is shown in Figure 11 .

Quality Control

Quality control of the freshly mixed repair material included: water/cement ratio, slump of the mix,sampling the mix for the compression tests, compression test results . The 28 day compression test result sranged from fc’=39 to 51 MPa, which exceeded the specified requirement of fc’=30 MPa .

The quality workmanship and the concrete properties of the placed repairs was verified by drilling 101 . 6mm diameter cores ‘in situ’ and performing pull-out tests as shown in Figure 12 . The pullout tests gaveaverage bond strength (repair material to old concrete) from 1 .2 MPa to 2 .1 MPa, which is above therevised value of 1 .0 MPa required by the Specifications .

47


Figure 12: Pull-out Test

The quality of the installation of the GFRP and compliance with the Specification was also tested b yperforming pull-out tests on four 101 .6 mm diameter cores . The pull-out tests gave average bond strengthof the GFRP to substrata (repair material or original concrete) from 1 .9 MPa to 3 .4 MPa. That was als oabove the value required by the Specification . The failure mode was never through the GFRP/substrat acontact joint, but ra

Figure 13 : Completed Repair South-West View

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Conclusions

Based on the results of the assessment studies and the repair work performed, it is concluded that :

1) The repair design and implementation of the Gentilly-1 ring beam has been successfull ycompleted .

2) The repaired structure is in good condition and is satisfying current functional, safety, design an daesthetic requirements . The life of the structure has been extended for at least the next fifty years .

3) The structure should be closely monitored. The behavior of the repair materials will be comparedwith the original design intent, and the required maintenance effort will be performed to ensurethat the design requirements characterized by service life are met .

Acknowledgement s

The authors would like to thank Vector Construction Group for completing the Project on time andsatisfying high quality requirements, and ISIS Canada for their technical support and expert opinion wit hthe development of the technical specification .

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50


The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures and Installation of anAutomated Cathodic Protection System .

LM Smith British Energy Generation (UK) LtdCA Hughes British Energy Generation (UK) LtdG Jones

C-Probe Ltd

Abstract

This paper reports on a project which carried out repairs to the headworks and associated jetty and marinestructures at Hunterston B nuclear power station . Although alternative sources of cooling are available, theheadworks and associated structures have important availability functions with regard to the provision o fcooling water for the nuclear power plant . These marine structures are situated in an exposed coasta llocation with consequentially aggressive environmental conditions . The jetty and headworks wereoriginally designed to codes that have now been superseded and in order to ensure that the repairedheadworks structure would have sufficient future service life an automated impressed current cathodi cprotection system was installed at the time of the repair works .

As a structural material concrete is strong in compression and weak in tension . In order that concrete maybe utilised in practical structures, reinforced concrete is provided with steel reinforcement that carries anytensile loads or stresses by composite action . It is therefore important that the steel reinforcement inreinforced concrete structures is maintained in good condition otherwise the capacity of the structures wil lbe degraded and reduced. The major threat to concrete structures is corrosion of the steel reinforcementand this is particularly the case in marine structures due to chloride contamination . The extent of corrosionmay be monitored by measurement of the corrosion potential of the steel reinforcement relative to areference electrode and cathodic protection may be employed .

The remedial works carried out to the jetty at Hunterston B are a typical example of the use of this type o frepair and monitoring . The Hunterston jetty is the only means of access to the station’s Cooling Waterintake headwork structure . The intake headwork is essential as it provides the sea water that is used as th ecooling water for the steam condensers within the power station . The approach jetty and the headworkwere constructed using conventional reinforced concrete in the late 1950s and the early 1970s respectively .

During routine structural inspections it was identified that the structures required remedial works primarilydue to the high degree of chloride contamination as a direct result of the environmental exposur econditions found at the site .

To ensure structural integrity over the remaining life span of the structures, an impressed current Cathodi cProtection (CP) system was selected for the headwork structure above Mean Low Water Springs (MLWS )level and a sacrificial anode system below this level . The impressed current CP (ICCP) system on thestructure above MLWS is divided into a number of anode zones and each zone is independently powere dand monitored. The anode for all the zones comprises a mixed metal oxide coated titanium mesh fixeddirectly to the repaired and prepared concrete substrate with proprietary fixing pins which hold the mes hrigidly against the concrete .

This paper describes the repair works carried out to the headworks and the function, technical details ,installation and performance of the cathodic protection system .

51


The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures and Installation of anAutomated Cathodic Protection System .

Introduction

This paper reports on a project which carried out repairs to the headworks and associated jetty and marinestructures at Hunterston B nuclear power station . Although alternative sources of cooling are available, theheadworks and associated structures have important availability functions with regard to the provision o fcooling water for the nuclear power plant . These marine structures are situated in an exposed coasta llocation with consequentially aggressive environmental conditions . The jetty and headworks wereoriginally designed to codes that have now been superseded and in order to ensure that the repairedheadworks structure would have sufficient future service life an automated impressed current cathodi cprotection system was installed at the time of the repair works . This paper describes the repair workscarried out to the headworks and the function, technical details, installation and performance of thecathodic protection system .

History

The remedial works carried out to the jetty at Hunterston B are a typical example of the use of this type o frepair and monitoring . The Hunterston jetty is the only means of access to the station’s Cooling Waterintake headwork structure . The intake headwork is essential as it provides the sea water that is used as th ecooling water for the steam condensers within the power station . The approach jetty and the headworkwere constructed using conventional reinforced concrete in the late 1950s and the early 1970s respectively .

The water intake head works structure is of reinforced concrete construction as shown in Figures 1 and 2 .During the mid-1980s repairs were carried out to the water intake headworks to remedy deterioratio nrelated to chloride ingress and reinforcement corrosion . These repairs were largely patch repairs but als oincluded :

Application of sprayed Gunite with an embedded secondary steel reinforcement mesh typically 125m min depth to the entire soffit of the upper deck slab .

• Application of black pitch extended epoxy resin to the upper deck slab soffit .

• Application of an epoxy paint system to the columns and intake shaft above high water level only .

No cathodic protection was included in the repair scheme .

By the mid-1990s the repaired areas and some additional areas of the structure had begun to deteriorat eonce more and, during routine structural inspections, it was identified that the structures again requiredremedial works . This was primarily due to the high degree of chloride contamination as a direct result o fthe environmental exposure conditions found at the site which was aggravated by the dynamic effect o fwave impact and storm damage on the .existing repair s

Repairs

As a structural material concrete is strong in compression and weak in tension . In order that concrete maybe utilised in practical structures, reinforced concrete is provided with steel reinforcement that carries any

52


tensile loads or stresses by composite action . It is therefore important that the steel reinforcement inreinforced concrete structures is maintained in good condition otherwise the capacity of the structures wil lbe degraded and reduced. The major threat to concrete structures is corrosion of the steel reinforcementand this is particularly the case in marine structures due to chloride contamination . The extent of corrosionmay be monitored by measurement of the corrosion potential of the steel reinforcement relative to areference electrode and cathodic protection may be employed .

To ensure structural integrity over the remaining life span of the structures, an impressed current Cathodi cProtection (CP) system was selected for the headwork structure above Mean Low Water Springs (MLWS )level and a sacrificial anode system below this level . The combined system therefore comprises :

Sacrificial anode system to all exposed surfaces between +0 .60m OD and –1 .325m OD, and

Titanium mesh impressed current anode and cementitious overlay to all surfaces above +0 .60m OD

The impressed current CP (ICCP) system on the structure above MLWS is divided into a number of anod ezones and each zone is independently powered and monitored . The anode for all the zones comprises amixed metal oxide coated titanium mesh fixed directly to the repaired and prepared concrete substrate withproprietary fixing pins which hold the mesh rigidly against the concrete .

The structure was divided into six Remedial Work Areas (RWAs) . A 40mm gap was left between th einstalled anode meshes in adjacent RWAs except in areas with double layer anode meshes where 50 mm o fthe top layer mesh was removed. All metallic objects within the protected areas were made electricall ycontinuous with the steel reinforcement . A 25 mm sprayed concrete overlay was generally applied over theanode mesh, which was increased to 75 mm at the top deck . In some areas existing concrete was cut backto maintain both external clearances and the overlay thickness .

The impressed current cathodic protection system chosen was the C-Probe Achilles system (Figure 3 )which allows the structure to be divided into a number of zones for both cathodic protection and corrosio nmonitoring by means of embedded electrodes . A zoned ICCP and monitoring system has superio rperformance to a single global system as the impressed current can be tailored to meet the requirements o fseparate areas of the structure .

Existing previous repairs to the deck slab soffit, including secondary steel mesh reinforcement, wer eremoved and the surface was profiled to maintain a minimum of 30 mm cover to the main reinforcement ,which was repaired as required. The sequence of removal and reinstatement operations (A through E) i sshown in Figure 4 . High pressure water jetting was used to remove the concrete and patch repairs (Plates 1& 2) . Where required, existing reinforcing bars that were excessively corroded were supplemented and/o rreplaced with new steel suitably anchored into the structure . Once the entire deck slab soffit had beenreprofiled and prepared a titanium anode mesh and anode ribbons were fixed and a sprayed concret eoverlay applied (Figures 5 & 6) .

Performance

The latest repairs have performed very well without deterioration over a period of six years and have notshown the rapid deterioration that occurred with previous repairs . This is considered to be due to theinclusion of the zoned ICCP and monitoring system and it is recommended that future repairs to importan tstructures in this type of location include such a system.

53


Figure 1

F ig u re 2

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Figure 3

55

Plate 1 Water jetting

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Plate 2 Exposed steel reinforcement after removal of cove r

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FEASIBILITY STUDY OF IE-SASW METHOD FOR THE NON-DESTRUCTIVE EVALUATIONOF CONTAINMENT BUILDING OF NUCLEAR POWER PLANT

Yong-Pyo Suh, Korea Electric Power Research Institute, Kore aJong-Rim Lee, Korea Electric Power Research Institute, Kore a

Jeong-Moon Seo, Korea Atomic Energy Research Institute, Kore a

ABSTRACT

The IE-SASW method that combines Impact-Echo (IE) method with spectral analysis of surfac ewaves (SASW) is proposed as a newly developed nondestructive testing method in concrete structures .This method is based upon the idea that IE method uses the elastic p-wave velocity measured from SAS Wmethod on the concrete member, and applied to specimens to evaluate its feasibility . It was shown that thethickness of the concrete structure member and the depth of the defects such as voids could be identifie dby IE-SASW method with good reliability . Additionally, the GPR (Ground Penetrating Radar) technique shave been applied to the same specimens in order to establish the performance and reliability of th eproposed method, and compared with IE-SASW method. The experimental studies show that it is mor epreferable to use the IE-SASW than GPR to detect the voids just beneath the steel reinforcing bars thatmay exist in concrete structures .

Keywords: Non-destructive Testing, Concrete, Impact-Echo, SAS W

1. INTRODUCTIO N

The construction quality of the containment building in the nuclear power plant is carefullycontrolled and thoroughly inspected to prevent from unexpected flaws . In general, the concrete ofcontainment building is deteriorating as time passes by, so the periodic safety assessments using non -destructive tests are also crucially required. Until now, the non-destructive tests such as radar, impact-echo, and ultrasonic methods have been developed for the concrete structure and compared thei rcharacteristics by various test specimens (M . Krause . et al ., 1997). Even though each non-destructivetesting method has its own advantage and capability, it is necessary for the user to understand thelimitations of the methods exactly before applying them . Also, the best combinations of testing method scan be selected only after comparative studies are performed .

In this study, IE-SASW testing method was introduced and the performance was evaluated byapplying to the containment building of nuclear power plant. The IE-SASW method combines IE (impact-echo) with SASW (spectral analysis of surface waves) methods, and is applied to detect the flaws as wel las the thickness of concrete structures . In the original IE method, the thickness and flaw depth in theconcrete structure are determined using the predetermined P-wave speed (Sansalone, 1997) . The accuracyof the result is much dependent on the accurate measurement of representative P-wave speed of the testin glocation. In the IE-SASW method, the representative P-wave speed is determined using SASW method .

In this paper, the performance of applying IE-SASW method to the non-destructive evaluation o fcontainment building of nuclear power plant was studied . Two test specimens were constructed and thedefects were included at the known locations. One of the specimens was the prototype of a structura lmember of the containment building typically built in Korea . IE-SASW method was performed to locatethe flaws and to determine the thickness .

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Additionally, the GPR (Ground Penetrating Radar) method is applied to the same specimens, an dcompared with IE-SASW method . The series of experimental studies were focused to detect the void jus tbeneath the steel reinforcing bars that may happen in concrete structures .

2. NON-DESTRUCTIVE TESTING METHODS USING ELASTIC WAVE

2.1. Impact-Echo Method

Impact-echo method is a nondestructive testing method of concrete structures that is based on thepropagation characteristics of impact-generated stress waves that are reflected by internal flaws andexternal surfaces. It can be used to determine the location and extent of flaws such as cracks ,delaminations, voids, honeycombing, and debonding in reinforced concrete structures (Sansalone, 1997) .A schematic diagram of impact-echo method is shown in Figure 1 .

The stress pulse generated by an impact on the surface propagates back and forth between th einternal interface and top surface of a test object. Surface displacements caused by reflections of thes ewaves are recorded by transducer (accelerometer) located adjacent to the impact . The resultingdisplacements versus time signals are transformed into the frequency domain, and plots of amplitud eversus frequency spectra are obtained. Multiple reflections of stress waves between the impact surface ,flaws, and/or other external surfaces give rise to transient resonance, which can be identified in thespectrum, and used to evaluate the integrity of the structure or to determine the location of flaws. Thetypical wave forms and amplitude spectrum of signal are illustrated in figure 2 .

The period of reflected waves is equal to the travel path 2T, divided by the compressive wave velocity, Vp .Since the frequency is the inverse of the period, the resonance frequency f, is :

f = Vp 2T

(1 )

This is a fundamental equation of impact-echo response for solid member, and if the Vp is pre-determined, the thickness and location of internal flaws can be identified .

The above analysis is valid for the cases where the reflecting boundary or internal interface ha slower acoustic impedance (density × P-wave speed) than the member . When this case occurs, P-waveincident upon the interface changes sign . For example, the P-wave generated by impact is a compressio nwave. When this wave is incident upon a solid/air interface, the reflected wave is a tension wave a sillustrated in figure 3(a) . If, however, a P-wave is incident upon an interface that has higher acousti cimpedance, such as steel reinforcing rebar in the concrete, or enlarged area, the incident P-wave does no tchange sign, and tension wave will be reflected at the interface (higher acoustic impedance) as a tensio nwave as illustrated in figure 3(b) . Thus equation (1) must be modified, because the period is twice as long .

f = Vp 14T

(2)

2.2. Spectral Analysis OfSurface Waves (SASW) Method

The spectral analysis of surface waves (SASW) method is a method of seismic testing that ha sdeveloped for determining shear wave velocity profiles at soil and/or pavement sites (Nazarian and Stoko eii, 1986) . The SASW method is a nondestructive method in which both the source and receivers ar elocated on the surface as shown in figure 4 .

The source is simply a transient vertical impact that generates a group of surface waves of variou sfrequencies in the medium . Two vertical receivers located on the surface monitor the propagation ofsurface wave energy . By analyzing the phase information of the cross power spectrum determine d

60


between the two receivers, surface wave velocity - wavelength relation is determined. If the stiffness of asite varies with depth, then the surface wave velocity will vary with wavelength . The variation of surfacewave velocity with frequency (wavelength) is called dispersion, and a plot of surface wave velocity versu swavelength is called a dispersion curve . The dispersion curve is developed from phase information of thecross power spectrum. This information provides the relative phase between two signals (two-channe lrecorder) at each frequency in the range of frequencies excited in the SASW test . In a homogeneou smedium, surface wave velocity, Vr , is constant and independent of the wavelength . Detailed procedure o fdetermining dispersion curve in the SASW method is described elsewhere (Joh, 1996, Nazarian and Stoko eii, 1986) .

In order to apply the SASW technique to the nondestructive test in concrete structures, the variatio nin surface wave velocity along the whole thickness of a concrete structure should be determined, and the P -wave velocity can be converted using Equation (3), assuming the Poisson's ratio of the concrete to be inthe range between 0.15 and 0 .2 .

V _

1 +V

2( 1 -v)V

3

p 0.87 + 1 . 12v (1 - 2v) R

( )

The probable error in determining P-wave velocity with the assumed Poisson's ratio of concrete betwee nzero and 0 .2 is about 3%, which is minimal, and with Poisson's ratio of 0 .2, the Equation (3) reduces to :

Vp = 1 .79VR

(4 )

2.3. IE-SASW method

In order to predict the thickness of concrete member or to identify the defect using IE method, P -wave velocity of concrete is required. The P-wave velocity can be found as presented in Equations (1) an d(2), when the boundary condition and the thickness of concrete member are predetermined . In general ,however, the thickness of member such as slab and wall of building remains unknown . The concretespecimen should be extracted from the structures using core-boring machine, and substituting the height o fcore specimen into the equations (1) produces the P-wave velocity. But this method has suchdisadvantages that the surface of structure is destructed due to core boring, and the P-wave velocit ycalculated may not be representative value of the structures because of the non-homogeneity of th econcrete material .

Therefore, IE-SASW method that enables to obtain the P-wave velocity from SASW methodnondestructively and then to apply the IE method, is proposed in this study . In order to obtain the materialproperties in the multi-stratified soil system using SASW method, the iterative inversion processing i srequired until the discrepancy between the experimental dispersion curve and theoretical dispersion curv eis minimal (Joh, 1996) . But in concrete structures, the surface wave velocity (Vr) can be easily obtained byexperimental dispersion curve without executing the inversion procedure, because concrete material i sassumed to be composed of a single layer . Using the relationships among surface wave velocity (V r), P-wave velocity (Vp), and Poisson's ratio ( v) with assumed Poisson's ratio of 0 .20, P-wave velocity o fconcrete can be determined by Equation (4) . Then, the thickness or the location of defect such as void inthe concrete member can be identified using IE method nondestructively .

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3. APPLICATIONS

3.1 Test Specimens

Two test specimens were made to study the feasibilities of IE-SASW method in the nondestructiv eevaluation of reinforced concrete structures . One test specimen, named “A” (length : 150cm, width: 50cm,thickness : 30cm, including two voids) was designed as illustrated in figure 5 . Voids were simulated withstyrofoam of which the acoustic impedance is distinguishably smaller than that of concrete . Especially, inthis specimen, one void is located just beneath the rebars and the other is located apart from the rebar. Thecover depth to the void beneath the steel reinforcing bar is 10cm and the other is 15cm. The mixproportion of the concrete is shown in table 1 .

In order to obtain the P-wave velocity, three concrete test molds were made and cured in a waterbath at 20 c for 3 days . After casting, the P-wave velocity was measured using IE method respectively ,and averaged to provide the P-wave of about 3300 m/sec .

Another test specimen, named “B”, is prototype structural member of the containment building of anuclear power plant with three tendon sheathing pipes (diameter of 150mm) that are unfilled and som edozens of rebars (diameter of 55mm) as shown in figure 6 . In order to compare the feasibilities of both IE-SASW and GPR methods, the cubic styrofoam (10cm x 10cm x 10cm) and water container (width : 20cm,height: 40cm, thickness : 40cm) filled with water, and a PVC pipe which is 15cm in diameter were inserte dintentionally during the construction at the depth of 30cm from the surface of the specimen . Thephotograph 1 shows the prototype member of the containment building of a nuclear power plant .

3.2 Results of IE-SASW method

3.2 .1 . Test specimen-A

The SASW test was performed on the test specimen-a to get the p-wave velocity before applying th eIE test . The interval between source and the first receiver is set to 20cm that is the same distance betweenreceivers . Figure 7 shows the dispersion curve produced by the phase difference between the tworeceivers, and the phase velocity is approximately chosen out to be 1870m/sec . The phase velocitybetween two receivers is equal to surface wave velocity without inversion process, because it can b eassumed that the surface wave velocity is constant and independent of wavelength in a concrete layer.Thus, the P-wave velocity can be calculated as 3347m/sec by the equation (4), assuming the Poisson’s rati oof concrete to be 0 .2 . It is resulted that the P-wave velocity of 3,300 m/sec obtained from the SASW tes tshows good agreement with the velocities that are previously obtained from the concrete test molds andfrom IE test using known thickness . Therefore, it is revealed that the p-wave velocity can be obtainedfrom the SASW test reliably.

The IE test was performed at the three positions : (1) on the void caged behind steel reinforcing bars ,(2) on the void where there is no rebar, and (3) on the surface where there is no defect through tes tspecimen. Several steel balls were used as impact sources, and accelerometer (PCB 353b15) was used as areceiver. Signals were recorded and analyzed using a dynamic signal analyzer (HP 35665a) . Theamplitude spectrums of acceleration at these three positions are presented in figures 8 (a), (b), and (c) ,respectively .

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The location of the void and the thickness of specimen can be determined by amplitud espectrums using P-wave velocity determined by SASW method, and substituting these resonanc efrequencies in figures 8 (a), (b), and (c) into equation (1) produces 10 .6cm, 15 .1cm, and 30 .1cm,respectively. Comparing these results with true values of 10 .0cm, 15 .0cm, and 30 .0cm, it can b econcluded that the IE-SASW method shows a good potential to identify the defect and unknown thicknes sof concrete member reliably .

3.2 .2 . Test Specimen-B

The P-wave velocity of test specimen was determined by the SASW test . A test was carried outalong the line where there were no defects and no rebars as designated in figure 6 . SASW test wa sperformed twice, changing the interval between source and the first receiver from 20cm to 40cm. Thedispersion curves are presented in figures 9 (a) and (b) . The P-wave velocities calculated from the surfac ewave velocities were 4200m/sec and 4100m/sec, respectively, and average value of 4150m/sec was used i nIE test.

A series of IE test were performed at the position (E1 to E13) designated in figure 6, and theamplitude spectrums of acceleration are shown in figure 10 . At position of E1, the resonance peakfrequency of 1824hz was observed, and the thickness of wall was calculated as 1 .14m, which is in goodagreement with the true thickness of 1 .2m. At the position of E5, E6, E11, and E12 where the metal sheat hpipes are located, the clear and large resonance peak frequencies were observed in the range from 4700h zto 4900hz. These frequencies produce the depth to the metal sheath pipe as 0 .45~0 .47 m, which fits wel lwith the actual depth of 0 .45m, and it is revealed that the metal sheath pipe have lighter impedance tha nconcrete and acts as a free boundary . However, the impact-echo response from the rebar of position E2shows the dominant peak at 4700hz, which does not give the actual depth to the steel reinforcing bar . Theacoustic impedance of rebar is about five times that of concrete, but the area of rebar is much less than thatof concrete . In this case, the reflected and refracted P-waves from the interfaces of concrete/steel o rsteel/concrete created by the rebar makes the signal to be complicated, because the location of rebar i sshallow but the depth to the bar is relatively deep . This fact makes it difficult to locate the rebar by IE test .

the amplitude spectrums corresponding to styrofoam, water container, and PVC pipe are shown in figures11 respectively . The impact-echo responses from these objects show large amplitude resonance betwee n6khz and 7khz. At the position of p1 in PVC pipe, the highest amplitude peak occurs at 6 .976khz: thispeak is used to calculate a depth of 0.297m, in good agreement with the known depth of 0 .3m. At theposition of s4 in styrofoam, the highest amplitude peak occurs at 6 .336khz: this peak is used to calculate adepth of 0.327m, which is close to the known depth of 0 .3m. Finally, at the position of w4 in watercontainer, the highest amplitude peak occurs at 6016khz : this peak is used to calculate a depth of 0 .345m,which is a little larger than the actual depth of 0 .3m. These results show that it is feasible to use the IE-SASW method to detect the voids (filled with either water or not) and PVc pipe .

2.4. Results of GPR method

2.4.1 . Test specimen-A

A ground penetrating radar (GPR) was employed and tested on both specimens A and B for th ecomparison purpose. Figure 12 shows the typical processed image corresponding to the profile of testspecimen-a, which recorded with 1200mhz antennas . This section shows reflected and diffracted signal s(hyperbolas designated by arrows) in response to the different objects : (1) six rebars, and (2) the voidwhere there is no rebar . Contrary to the ie method the void beneath reinforcing steel bars could not bedetected by GPR method.

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2.4.2. Test specimen-B

The GPR profile corresponding to the test specimen-b is shown in figure 13 . A series of rebarspacing evenly is illustrated as hyperbolas . However, it is difficult to identify the tendon sheath locate dbehind the rebar, because most of the electromagnetic wave is reflected when encountered with rebar .Pulse type dipole antenna of 1,200mhz (central frequency) failed to detect the objects such as PVC pipe ,styrofoam, and water container in the specimen-b .

3 .

Conclusions

The comparative studies presented here illustrate the performance and feasibility of nondestructiv etest to the concrete structures . The experiments were quite useful to select the most suitable techniques fo rspecific applications . In this study, the IE-SASW method and GPR method are employed to evaluate theirfeasibilities . The conclusions obtained are as follows :

(1) The IE-SASW method can be applied to identify the location of defect and the thickness ofconcrete structure with good accuracy .

(2) Especially, the void just beneath the rebar in test specimen-a could be easily detected by the IE-SASW method . On the other hand, the rebar itself could not identified by this method .

(3) The location of tendon sheathing and thickness of the structural member (test specimen-B) coul dbe identified by the IE-SASW method. Also the location of the objects such as styrofoam, watercontainer, and PVC pipe that were intentionally included in the test specimen-B can be detecte dby this method .

(4) The GPR method provides an objective and reliable image corresponding to the rebars an ddefects such as voids. But it was difficult to identify the void just beneath the reinforced stee lbars . Therefore, the detection potential can be improved by the combined utilization of the IE-SASW method and the GPR method .

REFERENCE S

[1] Carino, N. J ., and Sansalone, M., 1992, "Detection of Void in Grouted Ducts Using the Impact-Ech oMethod", ACI Material Journal Vol . 89, No . 3, pp.296-303 .

[2] Carino, N . J., Sansalone, M., and Hsu, N. N., 1986, "Point Source-Point Receiver Technique for FlawDetection in Concrete", ACI JOURNAL, Proceedings Vol . 83, No . 2, pp .199-208 .

[3] Joh, S. H., 1996, "Advances in Interpretation and Analysis Techniques for Spectral-Analysis-o fSurface-Waves (SASW) Measurement," Ph . D. Dissetation, The University of Texas at Austin.

[4] Lin, Y. and Sansalone, M ., 1992, "Detecting Flaws in Concrete Beams and Columns Using the Impact-Echo Method", ACI Materials Journal, Vol . 89, No . 4, pp .394-405 .

[5] Malhorta, V. M., and Carino, N . J., 1991, "HANDBOOK on NONDESTRUCTIVE TESTING ofCONCRETE", CRC Press, New York, 343p.

[6] M. Krause, M. Bormann, R. Frielinghaus, F . Kretzschmar, O. Kroggel, K. J. Langenberg, C.Maierhofer, W. Müller, J. Neisecke, M . Schickert, V. Schmitz, H. Wiggenhauser and F . Wollbold,1997, "Comparison of pulse-echo methods for testing concrete", NDT&E International, Vol . 30, No .4, pp . 195-204 .

[7] Nazarian, S ., and K.H. Stoke, II, 1986, In Situ Determination of Elastic Moduli of Pavements System sby Spectral Analysis of Surface Waves Method (Theoretical Aspects), Research Report Number 437 -2, U.S. Department of Transportation, Federal Highway Administration, pp .2-6 .

[8] Richart, F . E. Jr., Hall, J . R. Jr., and Woods, R. D., 1970, "Vibrations of Soil and Foundation", Prentic eHall, Inc ., Englewood Cliffs, New Jersey, 414p .

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[9] Sansalone, M ., and Streett, W . B., 1997, Impact-Echo Nondestructive Evaluation of Concrete an dMasonry, BULLBRIER PRESS, ITHACA, N .Y., pp .9-320 .

[10] Sansalone, M., and Carino, N. J., 1989, “Detecting Delaminations in Concrete Slabs with and withoutOverlays Using the Impact Echo Method”, ACI Material Journal, Vol . 83, No . 2, pp.175-184 .

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Figure 1 . Simplified diagram of impact-echo metho d

(b) amplitude spectrumFigure 2. Typical result of impact-echo test

f1 = VP / 2 T

f1Freq. f1

Freq.

void

(a) lower acoustic impedance

(b) higher acoustic impedanc eFigure 3 . Comparisons of resonance frequency between lower and higher acoustic impedance

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Styrofoam(depth=

Figure 4 . General configuration of SASW testing

Figure 5 . Drawing of test specimen-A

11 SPA . @ 35 .56 cm

rebar (diameter=55mm,

Figure 6 . Drawing of test specimen-B that is the prototype structural memberof containment building of nuclear power plant

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0 4000

Figure 7. SASW test result to determine the P-wave velocity of test specimen-A

0

10000

20000

30000

frequency(Hz)

(a) amplitude spectrum for the void behind steel reinforcing bars3 .0E-0 5

2 .0E-0 5

1 .0E-0 5

0 .0E+00

0

10000

20000

3000 0

frequency(Hz)

(b) amplitude spectrum for the void where there is no steel reinforcing bar1 .0E-05

0 .0E+00

0 10000

20000

3000 0

5 .0E-06

frequency(Hz)

(a) on the surface where there is no defect through test specimenFigure 8 . Amplitude spectrums of acceleration for test specimen-A at the three positions :

(a) on the void behind steel reinforcing bars, (b) on the void where there is no stee lreinforcing bar, and (c) on the surface where there is no defect through test specimen .

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0 4000 0

Phase Velocity (m/s)1000

2000

3000

Phase Velocity (m/s)1000

2000

3000 4000

Distance : 20 cmVR=2350 m/s

Poisson’ Ratio =0.2

Vp = 4200 m/s

0.00

0.60

0.80

(b) Distance between receivers : 40cm

0.00

0.30

(a) Distance between receivers : 20cm

VR=2300 m/s

Poisson’ Ratio =0.2

Vp = 4100 m/s

D istance : 40 cm

Figure 9 . SASW test results to determine the P-wave velocity of test specimen- B

rebar

E 4

E 5

E 6

E10

E1 1

E12

Figure 10. Amplitude spectrums of acceleration for test specimen-B at different positions (Theoretica lresonance frequencies corresponding to wall thickness and sheath location were plotted a sdotted lines respectively.)

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10000

20000

30000 0

10000

20000

Frequency (Hz) Frequency (Hz)

(a) Amplitude spectrum corresponding to PVC pipe (B) Amplitude spectrum corresponding to Styrofoam

10000

20000

Frequency (Hz)

(c) Amplitude spectrum corresponding to water containe r

Figure 11 . Amplitude spectrums of acceleration for PVC pipe, Styrofoam, and water container in tes tspecimen-B at different positions

(a) Schematic plan showing the location of reinforcing bar and void s

(b) Radar image of the reinforcing steel bars and voi d

Figure 12. GPR image at test specimen-A (scanned by 1200MHz antenna)

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profile

Reinforcing steel bar

Metal sheath

(a) Schematic plan showing the location of profile with reinforcing steel bars and metal sheath s

Figure 13 . GPR profile at test specimen-B (scanned by 1200MHz antenna )

Table 1 . Mixture proportion of test specimen concrete

Water Cement Aggregate oS/A(/o) Air Content(%)

W/C(%)Coarse Fine

185 320 1026 713 41 5 58

Photograph 1 . Prototype Structural Member of Containment Building Of Nuclear Power Plan t

7 1


72


Field Studies of Effectiveness of Concrete Repairs

NJR Baldwin, Mott MacDonald, UK

Abstract

This paper presents a summary of the work carried out under an HSE research study entitled `Field Studie sof Effectiveness of Concrete Repairs' .

There is little published information describing or comparing the long-term performance of different repai rtypes. This project examines 45 sites where repairs have been carried out and have remained in service fo rseveral years and evaluates the effectiveness of a range of concrete repair systems as applied in practice .The sites include a range of structure types, ages and service environments, and include bridges, tunnels ,building frames, and car parks in industrial, public, highway and nuclear environments . The repairsinclude hand and trowel applied materials, sprayed materials, and cathodic protection techniques, and als osome sites with coating and crack injection systems .

The objective is to improve practices for maintaining and improving the integrity of operational structure sand so achieve higher standards of structural safety and reliability and better whole-life structuralmanagement .

The project employed a range of visual, non-destructive and destructive investigation techniques atrepaired sites and compared the condition found with records of the repair procedures and objectives . Thesite investigations involved detailed visual inspection, and surveys using hammer tapping, covermeter ,half-cell and carbonation depth, pull-off testing and core sampling for petrographic analysis of the repai rmaterial, repair layer interfaces and repair/substrate interface .

Examination of records of the repairs allowed assessments to be made of the level of understanding of theoriginal cause of deterioration and the need and objectives for the subsequent repairs . Comparisons havebeen made between the specification and evidence from the repair sites . The owner's repair objectives andconstraints, and the quality and effectiveness of the repairs have been considered .

The site investigations have provided detailed information on the performance, structure and effectivenes sof repairs . Full records of the repairs were characteristically difficult to obtain . Most repairs were tocorroded reinforcement in structures affected by chloride-induced corrosion . There was often evidence ofpost-repair corrosion with the exception of those structures with a cathodic protection system . Fine cracksand surface crazing were common in the repair materials and at the perimeter of patch repairs . These canpenetrate to the reinforcement and represent performance limiting features .

The project output includes production of guidelines covering the decision making process of concret erepair. These guidelines will be disseminated to industry via the HSE and through new guidance noteswithin industry documents and a proposed ICE publication .

1 .

Introduction

1 .1

Origin

Mott MacDonald Ltd (MM) was commissioned by the UK's Health and Safety Executive (HSE) in June2000 to carry out a research study entitled `Field Studies of Effectiveness of Concrete Repairs' . The projectfollows on from the project conducted by the UK's nuclear industry entitled `Concrete repair materials an dprotective coating systems' that produced a compendium of repair materials and systems 1 .

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In 1999 the effectiveness of in situ repairs was added to the HSE’s Nuclear Research Index . The issuequestioned the potential performance, maintainability and longevity of nuclear safety related structure swith extended service and safe storage lives using current concrete repair practices . HSE’s NuclearIndustry Inspectorate (NII) proposed a programme of field studies to examine typical repair sites . NIIrecognised the difficulties in gaining access to repair sites within nuclear facilities and in 2000 secure dfunding from the HSE to conduct field studies on the repairs to a variety of concrete structures away fromnuclear sites .

The scope and objectives have been developed between the HSE, MM and other organisations whos einterests are represented in an Expert Group associated with the project . Funding has also been receivedfrom the Highways Agency (HA) and the Institution of Civil Engineers (ICE) Research and DevelopmentEnabling Fund . The project receives substantial additional support from collaborating organisation andindividuals, as well as the co-operation of owners of repaired structures .

1 .2

Objectives

The aim of this project is to evaluate the effectiveness of a range of concrete repair systems as applied inpractice, in order to improve practices for maintaining and improving the integrity of operational structure sand so achieve higher standards of structural safety and reliability and better whole-life structuralmanagement. This includes assessment of the whole process whereby repair is carried out, and i nparticular what parts of the process lead to success or failure. The project also investigates the effects o fageing of repairs and identifies the most effective means of providing enhanced durability . Guidelinescovering the decision making process of concrete repair will be disseminated to industry .

The project has focussed primarily on non-structural patch repairs to reinforced concrete intended to arres tdeterioration resulting from the ageing processes prior to any significant effect on structural integrity suc hthat structural intervention is avoided . Structural repairs, in which the load paths pass through the repair,have not been specifically targeted .

Investigations have also been carried out into the long-term performance of practical cathodic protectio n(CP) schemes. The performance of the systems has been assessed, and the effects of CP on the bond ofreinforcement and the distribution of ions throughout the concrete has been investigated .

1.3 Definitions

The terms used within the study have been the subject of intense debate . The problems lie in findingdefinitions which are agreeable for civil, structural and materials engineers, the repair industry an doperators of structures, for terms such as defect, repair, effectiveness, performance, non-structural repai rand structural repair. Examples of definitions from ENV 1504-9 2 and EU FP5 research project“LIFECON” are presented in Table 2 .

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Table 1 Definitions

ENV 1504-9 EU FP5

Defect An unacceptable condition whichmay be in-built or may be the resul tof deterioration or damage .

Repair A measure which corrects defects . Return of a structure to an acceptabl econdition by the renewal, replacementor mending of worn, damaged ordegraded parts .

Protection A measure which prevents or reducesthe development of defects .

The definition and objective of repair works are also variable and the following definitions are found in theliterature :

Etebar3 defines the objective of repair as being to restore or enhance one property such as durability ,structural strength, function or appearance .

Walker4 states that rehabilitation involves controlling degradation to enable a structure to continue t oserve its intended purpose, either through repair to a state similar to the original, or using methods toarrest deterioration processes .

Emmons and Vaysburd5 , state that the object of any repair project is to “produce a repair at relativelylow cost with a limited and predictable degree of change over time and without deterioration and/ordistress throughout its intended life andpurpose”.

These demonstrate the different elements and concepts of the objectives of repair and intended repai rperformance. Measurement of repair effectiveness is more complex and subjective, and is best addressedby comparing condition over time relative to the original objectives .

1.4 Effectiveness

A repair may be effective if it has achieved the performance that it was originally intended to . However,there may be several aspects to the original intention, such as cost, longevity and cosmetic issues . Theremay also have been requirements or restrictions for preparation and application of the repair, that form par tof it’s effectiveness . This project has used a detailed definition of effectiveness, adopting the principles ofthe ‘SPALL’ criteria, defined by King and Ecob6 and listed below .Structural : Possesses the required structural properties .

Protection : Provides protection for the reinforcement.

Application : Can be applied effectively within the given constraints .

Longevity: Once applied it remains in place .

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Looks :

Has an appropriate surface finish .

The ‘SPALL’ parameters, in combination with assessment of the quality of the processes of selection an dspecification of the repair, have been reviewed at individual sites . Site observations and testing have beenused to gather information on the effectiveness of the technical, site activity aspects of the repair, i .e . thephysical actions involved in execution of the repair . The contract information and records have been use dto evaluate the process effectiveness, i.e . the planning and management of the repair.

1 .5

Structure

The project was carried out in four main stages . The first was a data gathering phase in which the literatureavailable for the types, methods, standards, investigation techniques and performance of repairs wa sreviewed7 . This confirmed there is little independent data on the long-term in situ performance of mostconcrete repair systems . The second stage involved identifying the sites to be visited, contacting th eowners/operators, sourcing record documents, and executing risk assessments and designing th einvestigation procedures . The third stage involved visiting 45 sites to form a database of defects 8 .Specialist techniques were used to investigate a large number of repairs and CP systems at one major site 9.The final and current stage of the project involves the analysis and presentation of the data 10 . All of thereports will be available via the HSE in 2002 .

2 .

Site investigation

2.1 Number and type of sites investigated

Patch repairs were targeted where the depth of repair was less than the full depth of the element, typicallyless than 100mm deep, and the areas repaired were typically less than 1m x 1m. A wide variety of repairsand repaired structures were included to provide a representative population for study . In total, 45 site swere visited, and have been described in detail in relation to age, natural and service environment, repai rhistory and condition. However, the sites remain anonymous . At certain sites, there was more than onetype, generation or condition of repair, and in total 65 locations were examined .

The sites were located in throughout England and included infrastructure, public buildings and nuclea rfacilities in coastal, estuarine, river and inland locations . Repaired elements included beams, columns ,slabs and walls from structures including bridges, tunnels, power stations and other reinforced concret eframe buildings . The type and environment of structures visited is summarised in Table 2 and Table 3 .

Table 2 Types of repair examined

Structuretype andlocation

Concrete frame structure Carpark

Road bridge/viaduct Roadtunnel

Others

Public Industrial Inland Inland River Estuarine Estuary InlandunderpassInland Inland Estuarine Coasta l

Total 2 2 4 6 5 14 2 4 5 1

Total 14 5 20 5 1

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Table 3 Types of repair examined

Type ofrepair

Hand/trowel applied Sprayed Flowable

CP CrackSealingproprietary conventional proprietary conventional

dense high buildTotal 25 8 7 9 3 9 5 3

Total 40 12 9 5 3

At 48 locations the repairs were investigated through visual inspection, hammer surveying, non-destructivetesting, and intrusive sampling, and 60 core samples were examined in the laboratory . At 13 locations therepairs were examined by visual inspection and hammer surveying . At 3 further locations visualinspection was supplemented with non-destructive testing (NDT) .

The sites were selected where safe access could be achieved to repairs where some knowledge o rdocumentation of repair age, locations, type and method existed . The repair locations were selected torepresent a range of sizes, types and application methods . The repairs were applied by hand or trowel ,spray, or flowable methods . Mostly the repairs were of cement-based conventional materials or proprietar yrepair systems, and up to 12 years old. Seven different CP systems were also examined at four differentstructures. Each was associated with repair with cementitious materials prior to installation of the C Psystem.

2.2

Records of repai r

For each site, an attempt was made to discover the maintenance history, the cause of deterioration, and th eparties involved in the repair process. This involved consultation with the structure owner/operator andrecovery of documents related to condition prior to repair, the cause of defects, the repair contract ,specification and method statement, and detailed description and/or data sheets for the materials used in th erepair . Photographs of the repair contract were also invaluable.

The level of documentation available varied greatly. The older the repairs, the more difficult it was t orecover all of the information. There was also a difficulty in recovering documents where the ownership o fthe structure had changed . Where more than one repair generation occurred, there was often n oinformation relating to the earlier repair phases .

For a number of sites no records were available, or records of repair did not appear in existing engineerin gand maintenance files . Records from Health and Safety files did not contribute significantly to th einformation recovered for the sites, despite a requirement under CDM regulations 11 for retention of theHealth and Safety file for the life of the structure .

2.3 Road Tunnel Deck Study

The research also took advantage of replacement of the reinforced concrete deck of a major UK roadtunnel during 2000-2001 . Sections of the deck containing repairs and CP systems were inspected an dsampled. A total of 26 cores and 18 sections measuring approximately 600mm x 500mm, were sawnthrough the full depth of the deck . The sections were water jetted to expose a 250mm length of each bar i none face, and pull-out testing was carried out on 57 bars at the University of Birmingham .

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3 .

Site investigation methods

The performance of the repairs was assessed through a combination of visual inspection, non-destructivetesting e .g. hammer tapping, half-cell potentials, cover meter survey, carbonation depth . Destructivesampling and testing was carried out where permitted by the structure owner, and included pull-offresistance, extraction of a core for petrographic examination and electron microscopy, assessment o fcarbonation depth, and sampling for chloride and alkali contents . Other investigation techniques, such a schloride and alkali determinations were used at specific sites where deemed necessary . The sample hole swere repaired with a proprietary high-build repair system and the site cleared and vacated .

Where possible, more than one repair was sampled, identifying contrasts in condition, appearance, age ,location, and exposure conditions .

3.1 Effectiveness of investigations

Visual inspection and hammer tapping were of greatest value on site, particularly when combined withcore sampling or break out . The visual inspection provided a rapid assessment of the overall condition o fthe structure, its environment, and condition of repairs and coatings . Inspection of defects, particularlywhen combined with documentation of the condition and maintenance history of the structure, provided aninsight into the causes, severity and timing of deterioration . In comparison, the value of the half-cell surveyand pull-off resistance tests were limited .

3.2 Half-cell surveys

Half-cell surveys were carried out at many locations and most required careful interpretation . In general,there was a marked difference in potential values for the repairs and for the substrate concrete . Typicallythe repairs had a less negative value, and was often positive, indicating a dry substrate . A common trendfound in columns and walls, in both repairs and substrate, was of increasing negative potential toward sground level, and particularly within 0 .5m of ground level . This is associated with the increase in moisturecontent of the substrate and not necessarily associated with corrosion activity . However, in the road andcar park structures, chloride concentrations may be higher in the lower portions of elements exposed to sal tspray . This can result in higher corrosion potentials . The portions within 1 .5m of ground level exposed tospray were observed to have a high incidence of deterioration and repair .

Where reinforcement was exposed at the surface, either through low cover or through spalling of the cove rconcrete, an increase in negative potential was typically found only by detailed mapping and often over avery limited area. This was probably related to the general lack of moisture within the dense repai rmaterials . Consistently high negative potentials were found only in wet substrates, for example beneat hleaking joints .

Previous research 12 has found the surface zone of cementitious systems to have significant effects onoxygen diffusion and resistivity. These ‘skin effects’ can interfere with half-cell surveys and provide aninaccurate impression of corrosion potential at depth .

3.3 Pull-off resistance

The pull-off testing was characteristically problematic and time consuming particularly on overhea dlocations and on rough substrates . Failure typically occurred at the interface between repair material an dthe substrate (36%), within the concrete substrate (21%) or in the adhesive (25%) . The recorded failurevalues typically ranged between 0 .1 and 0.9N/mm2, with a mean value was 0 .48 N/mm2 and standarddeviation of 0 .22 N/mm2 .

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No difference was found between pull-off resistance in repairs with substrates prepared by waterjettingcompared to those prepared by mechanical break out . However, other research has demonstratedsignificantly higher bond strength on overlays applied to water jetted substrates compared with san dblasted and mechanically broken substrates 13 .

3.4 Core sampling and petrographic examinatio n

Examination of the core samples and core holes provided information on macroscopic feature such as voi dcontent and distribution, discontinuities, cracks, condition of reinforcement, and configuration of repairsand interfaces. The petrographic examination of the repair and substrate provided detailed information o nthe microstructure and quality of the repair material, substrate and the interface between them . Thisidentified differences and subtleties in repair composition and structure that were not appreciated on site,such as the extent and depth of fine cracking and microcracking from the external surface, the depth ofcarbonation along these features, layering in repairs and subtle differences in composition of the binder .

4 .

Preliminary findings

4.1 Reason for repair

The majority of the repairs examined related to corrosion of the reinforcement embedded in the originalconcrete substrate . Overall, approximately 60% of the locations examined had deteriorated, or wer erepaired, as a result of chloride induced corrosion or the potential for it. Approximately 25% haddeteriorated as a result of carbonation induced corrosion, and the remainder by drying shrinkage or othermechanisms. The structures built between the 1930’s and 1950’s had mostly required repair as a result o fcarbonation induced corrosion . These were almost exclusively in locations devoid of an external source o fchlorides . In most structures there was evidence of some post-repair deterioration. This was oftenassociated with the original cause of deterioration, such as chloride ingress . Performance of the repair swas not always a measure of the effectiveness of the management strategy .

4.2 Repair structure

Many features of the repairs could be identified, such as sawn perimeters, feathered edges, differentmethods and depths of break-out, presence of bond coats, reinforcement primers, levelling coats, externalcoatings, and the occurrence of layers, partings, cracks and voids . The components and proceduresdescribed in the repair records did not always match those found in the site investigation . The thickness ofapplied layers in hand or trowel applied materials was often greater than that recommended in the materialsdata sheets . There was little evidence that this had compromised repair effectiveness .

Many of the repairs were finished flush with the surrounding surfaces, resulting in reinstatement of lo wcover depths, sometimes less than 10mm . Repairs at several locations had an intentional local increase incover to overcome this . At other sites, the cover was enhanced by application of cementitious render oradditional protection was afforded by a paint or high performance coating at the external surface .

4.3 Cracking in repairs

Cracking occurred at the external surface in approximately 60% of repair locations examined.Approximately 45% of repair locations contained surface crazing and/or cracks resulting from dryin gshrinkage . The cracks commonly passed from the external surface of the repair to the embeddedreinforcing steel, and can pass through the full thickness of the repair . The drying shrinkage cracks aretypically less than 0.1mm wide, and the fine cracks grade into microcracks, clearly visible in the repair sunder petrographic examination, and remain readily resolved at 0 .005mm width.

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There is evidence of penetration of water into cracks and carbonation of the binder adjacent to cracks . Atapproximately 20% of repair locations the depth of carbonation along the cracks was greater than theminimum cover depth at the site or the depth of reinforcement intersected in the core sample . Fine cracksand microcracks in repairs less than a year old were carbonated to depths of up to 21mm . The depth ofcarbonation from the crack surfaces was very limited and typically no more than 2-5 times the width of thecrack .

Cracks were also found at the margins of repairs . The extent of carbonation in the repair material wa smostly insignificant, but the substrate concrete was commonly more porous than the repair, and carbonate dto greater depths . This is significant where reinforcement bars pass from the substrate into the repair andlocal depassivation at the steel could result in corrosion . The ingress of chlorides into cracks is also knownto result in corrosion of reinforcement at the perimeter of repairs 14 .

4.4 Deterioration of repairs

Deterioration had occurred to some ‘holding’ repairs in aggressive environments . A small proportion of‘long-term’ repairs were failing prematurely. The cause of deterioration was mostly site-specific andincluded reoccurrence of the original cause of deterioration, ingress of water and chloride, shortcomings inthe specification, crazing or cracking in the repair material, carbonation of poor quality material, vibratio nin the structure, reinstatement of low cover and the presence of a cavity at the repair/substrate interface.Repairs were ineffective mostly in the ‘Protection’, ‘Longevity’ and ‘Looks’ aspects of the SPALL criteria .

Evidence for incipient anode formation was found at several sites with cracking or spalling of the repairand/or surrounding concrete, typically at or close to the repair perimeter . All of these locations had beenrepaired because of chloride-induced corrosion . There were also locations where corrosion had continuedat the repair site, not necessarily through incipient anode formation, but mostly through ongoing exposur eto saline water, or the presence of carbonated concrete adjacent to the repair . The repair planning andmanagement processes for these sites were not fully effective .

4.5 Performance of CP

Each of the sites where CP had been operational showed very low levels of deterioration to the substrat econcrete and patch repairs . Some deterioration of the external coating systems was noted. The mean bondstress determined through pull-out testing on 57 bars was 3N/mm 2. The pull-out testing found no evidenc ethat CP had affected the bond strength for plain round bars in original deck concrete or the sprayedconcrete repairs in areas with CP and in areas without CP .

Examination of core samples confirmed that there was no evidence of significant corrosion of the bar sembedded in concrete protected with CP systems . Ionic mapping of the binder in repairs and origina lconcrete confirmed trends of ionic concentrations at the surface anodes and embedded cathodic stee lreinforcement. No evidence was found for significant deterioration resulting from these concentrations .

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5.

Preliminary Conclusions

This project aims to define and assess the effectiveness of in situ patch repairs and CP systems. A largepopulation of repairs of up to 12 years age have been investigated .

Records for the repair contracts have been examined but are not always complete or available .

The majority of structures had repairs to corroding reinforcement, caused predominantly by chlorid econtamination . Many of the repaired structures showed evidence of post-repair deterioration, particularl ywhere the cause of deterioration was not effectively treated. However, the structures with CP systemswere in notably good condition .

Many of the patch repairs contain shrinkage cracks which represent a performance limiting factor .

6. Acknowledgements

The author acknowledges the following funders of the research :

Health and Safety Executive .

Highways Agency .

Industry Management Committee .

Institution of Civil Engineers Research and Development Enabling Fund .

MM Group Research and Development Fund.

7.

References

1 Sheffield University Centre for Cement and Concrete, Mott MacDonald Special Services Division, DEW-Pitchmastic PLC (1999), "Concrete repair materials and protective coating systems", Nuclear repair contractBL/G/31221/S, IMC Reference CE/GNSR/5020 .

2 ENV 1504 (1997), "Products and Systems for the Protection and Repair of Concrete Structures - Definitions,requirements, quality control, evaluation of conformity", Part 9, "General Principles for the Use of Products andSystems "

3 Etebar, K., "Integrity of repaired concrete under repeated loading conditions" pp 493-502 Proceedings of theinternational conference on concrete repair , rehabilitation and protection, University of Dundee, Scotland, UK ,27-28 June 199 6

4 Walker, M., "An overview of rehabilitation methods and selection of an appropriate system" . Pp169-180.Proceedings of the international seminar on controlling concrete degradation, University of Dundee, Scotland ,UK, 7 September 199 9

5 Emmons, P .H. & Vaysburd, A.M. (1993), "Factors affecting durability of concrete repair - The contractor' sviewpoint", Proceedings of 5th international conference on structural faults and repair, Vol.2 Extending the Life ofCivil & Building Structures, University of Edinburgh .

6 King, E .S . & Ecob, C.R. (1993), "Review and Specification of Concrete Repair Materials" .

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7 Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R109 9‘Phase 1 Report’, April 2001 .

8 Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R112 0‘Phase 3 Report : Records of site investigation’, 2002 .

9 Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R112 1‘Phase 3a Report: Investigations at UK Tunnel’, 2002 .

10 Mott MacDonald Report for HSE Research ‘Field Studies of Effectiveness of Concrete Repairs’ : Report R113 4‘Phase 4 Report’, 2002

11 Construction (design and management) Regulations, 1994 (amended 2000), UK Legislation, Stationary Office Ltd,London .

12 M B Leeming, Concrete in the Oceans – Phase II – Co-ordinating report on the whole programme, Fina lReport to contributors, July 1986 .

13 J Silfebrand ‘Improving Concrete Bond in Repaired Bridge Decks’, Concrete International, September 199 0

14 TRRL Contractor Report 209, J G Keer and J R Chadwick, ‘Corrosion of reinforcement embedded in concreterepair materials exposed to de-icing salts’, 1990.

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OECD-NEA WG 10-11 .04.2002 BERLIN

Detect and repair of defects on th econfinement structure at Paks NPP

CSABA NYARADI

system technologist engineer

Paks Nuclear Power Plant Ltd .

P.O.B 71 . 7031 Paks ,[email protected]

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ABSTRACT

Paks Nuclear Power Plant is the only commercial nuclea rfacility in Hungary, which has been operational since 1982 .Like other N-plants, Paks NPP is also exposed to majo rcha llenges due to plant aging and changes incircumstances that affect the operation . T he defense indepth concept is a corner of nuclear safety, therefore th econfinement concrete structure and its aging phenomena ison the focus of the utilities . Declaring the lifetime extensionprogram, Paks NPP pays distinguished attention onconfinement integrity and achieved significant results on thisfield .

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Content of presentatio n

1. Introduction2. The construction and structure of hermetic spac e

(confinement.)3. Methods of examination of leakage and strengt h

test4. The results of leakage tests .5. Repairing of defects found during examinations .6. Short review of repair of the liner and concret e

structure at the Unit 1 . of NPP Paks .7. Conclusions.

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1 . Introduction

The third element of the defence in depth concept is th econfinement which provides the last physical barrie ragainst activity release to the environment .

Therefore, the leakage of the confinement is strictl ylimited, which is one of the fundamental conditions o fplant operation .

To assess the integrity of the confinement, the utility ha sto determine the leak rate of the confinement, even therehas been a locally uncontrollable hole opened on the liner .Outage for maintenance and refuelling is a typical case .

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2. The construction and structure of confinement

Net volume of confinement is 50 000 ÷ 54 000 m3 .Lower stage is –6 .5 m-, higher stage is +40 .5 m.

Carbon steel liner is on the inner side of confinement’s rooms ,and outer side of reactor compartment walls .Steel liner’s thickness is 8 mm.Thickness of concrete is 800÷1500 mm .

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Pressure transient at a LB LOCA DBA at VVER-440 confinement .

A [16 . s] - HPSI startedB [30 . s] - Water spill

back from the tray sstarted

C [48 . s] - Water spillback from the tray sfinished (90 % of watervolume spilled back)

D [86 . s] - spray injectio nstarted

Containment pressure(492 mm - double ended cold leg

50

100

150 200

250

300

Time [s]

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inentConfinement

n building i with the

b l Bubble Condenser

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3-D modell of confinemet

90


Comfinement’s 3-D modell

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3. Methods of examination of leakage and strength test

We had carried out a leakage test in full designed presssure -250 kPa - at first start-up phase. We had measured the leakageof confinement at three gaugh level ;120, 170 and 250 kPa .We had measured deformation of walls at same time .

Repeted leakage tests are achieved on 120 kPa .Leakage rate is converted to 24 hours and 250 kPa .We had managed a leakage test at 170 kPa on each unit a t1994÷1997 .We use the two point method, while pressure drops .

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Leakage rate of NPP units 1÷4.

1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1994 1995 1995 1996 1996 1997 1997 1998 1999 2000 2001 200 1

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5. Repairing and checking defects found during examination s

Defects of kiner- welding --> local test- injection plastic material --> integrated leakage tes t

Defects of hermetical closure s- exchanging gasket material --> local tes t- exchange sealing equipments --> integrated leakage tes t- making new gasket construction --> integrated leakage tes t- resealing of door’honges --> integrated leakage tes t

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6. Short review of repair of the liner and concret estructure at the Unit 1 . of NPP Paks

We have made a program improve hermetization of unit 1 .from 1998 co-operating with company VUEZ a.s . Slovakia.Mean activities :- searching defaults during ILRT and depression tes t- injecting the space between liner and concret e- measuring alteration of pressure gaugh under liner whilecarried out an ILRT- decomposition of concrete to liner at blowdown air corrido rand injecting- decompositioning of concrete at a hermetic roo mexplorating a connection to liner and repairing i t

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Injektion point No 1 ./19 deconstructed concrete in A201/ 1

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Injektion point No 1 ./19 with injection cap

97


Injektion point No 1 ./19 with local control cap

98


Excaveted connection to liner in room A306/ 1

99


Excaveted connection to liner in room A306/ 1

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Repaired connection in room A306/ 1

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Excaveted connection in room A306/ 1

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Repaired connection in room A306/1

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Hole in structural concrete discovered by drilling for injektio n

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Inside of hole in structural concretediscovered by drilling for injektio n

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Checking hole No 1 .for borical leakage and concrete at unit 2

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Checking hole No2 . for borical leakage and concrete at unit 2

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108



109



110


Checking hole No 3 . - unit 3

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Other activities to repair estate of concrete :

- gap of concrete between the spent fuel storage pool an dNo 1 . pool

- destructive and chemical disquisition of concrete sample s

- model supported mechanical and chemical experiments havebeen performed

- long term program to check the status of concrete structur e

- evaluation of concrete structure for lifetime extension

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7. Conclusions

7.1 . Notable degradation of concrete hardness has not beenfound.

7.2. Surface corrosion has been found on the metal structure i nthe concrete

7.3 . The confinement leak rate is within the limits of Techn.Spec.

7.4. Leak tightness enhancement program on the unit 1confinement is in progress, significant results are achieved .

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114


SESSION A: OPERATIONAL EXPERIENCE (continued )Chairman: Dr. James Costello, USNRC (USA)

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116


Steam Generator Replacement At Ringhals 3Containment, Transport Opening

Jan Gustavsson, Ringhals Nuclear Power Plant, Swede n

Abstracts

At the steam generator replacement at Ringhals 3 1995 an opening 6 x 8 m was taken in the cylindricalcontainment wall about 12 m over the ground level. The wall is build up of an outer prestressed concret ewall, steel liner and an inner concrete wall . The prestressing reinforcement in the outer concrete wal lconsists of horizontal and vertical tendons . Each tendon was prestressed to about 5000 kN from thebeginning . The tendon ducts were filled with grease.

Before the opening was taken a temporary wall with a gate was build on the inside of the containment wall .When starting making the opening, the tendons within the opening were de-tensioned and pulled out . Aadditional number of horizontal tendons under and over the opening were de-tensioned . The opening wascut through drilling 300 mm holes along the side of the opening into the steel plate from both outside an dinside . Then the steel plate was cut and the wall plug removed .

During the transport of the steam generators the work with the restoring of the wall started . The concretesurface was prepared and reinforcements bars were drilled into the old concrete . When the transport wa sfinished the steel plate was restored, tendon ducts joined and the reinforcement put in place. To preventearly cracking in the concrete tubes for cooling water were mounted in the form . The inner form wascasted at first and then the outer one .

After that the concrete has reached sufficient strength the tendons were mounted in the ducts and tensione dto the same force as before de-tensioning. After the tensioning the ducts were filled up with grease again .

A special inspection program was performed to see if there were any degradation in the concrete, thetendons or the steel plate . The result was that we could see no degradation on the steel plate or the tendons .The measured tendon forces are higher than calculated. The strength of the concrete had reached a mediu mlevel of 91 MPa (Original level 50 MPa) . The carbonisation had reached 8 – 10 mm into the concrete . Ourconclusions are that we have not seen any degradation which is a threat to the containment in th eforeseeable time.

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1 .

Introduction

1.1 Background

In 1992 Ringhals decided to change steam generators at Ringhals 3. The degradation of the tubes had gonefar and the capacity was reduced to 88 % since 1988 . A steam generator replacement had been done a tRinghals 2 during 1989 and the experiences from the replacement were good .

The planing of the replacement started during 1992, and KWU was chosen to manufacture the stea mgenerators . For the replacement at Ringhals a consortium consisting of Siemens and Framatome wa sengaged. The civil works were carried out of a subcontractor, NCC, to the consortium . NCC is one of thebiggest contractors in the Swedish building industry . The replacement work started in the beginning o fJune 1995 and was finished within 90 days .

1.2 Transport procedure

The steam generators were transported to Ringhals by boat to the harbour at Videbergshamn near RinghalsSite . From the harbour the steam generators were transported by a special vehicle to a storing place withi nthe Ringhals area . Some preparation works were done on the steam generators . The replacement procedurestarted with making the opening. As soon as the opening were ready the old steam generators were takenout from the containment one at a time. The steam generator was lifted up from its position by the polarcrane and laid down on trolleys on girders at level +115 . Then the steam generator was transported throughthe containment opening and lifted down to ground level by a lifting device . The vehicle then took thesteam generator to a storing place . When the three old steam generators were taken out it was time for thenew ones to be transported into the containment. The transport procedure was the same in the opposit edirection .

1.3 Description of the wall

The cylindrical containment wall consists of one outer concrete wall, a steel liner and an inner concret ewall . The inner diameter of the containment is 35 .4 m and the height is 52 m .

The inner concrete wall is 0 .33 m thick concrete reinforced with a grid of 016 mm rebars distance 200 m min the inside. The compressive strength of the concrete in the wall is K50 .

The steel liner is 7 mm thick . On the outside of each joint between plates in the steel liner there is U-profil ewelded in order to make it possible to see if there is any leakage through the joint. The canal inside the U-profile is connected to each other in special sections . From each section there is a pipe drawn to a galler ywhere it is possible to test each section . To the steel liner there are profiles welded to which the tendonducts are fixed .

The outer concrete wall is 0 .77 m thick concrete reinforced with a grid of 0 25 mm rebars distance 200mm in the outside . The covering concrete layer is 50 mm. The compressive strength of the concrete is K50 .About 0 .3 m in the wall there are horizontal tendons and about 0 .5 m in the wall there are vertical tendons .The distance between the vertical tendons is 0 .75 m and the average distance between the horizontaltendons 0.4 m .

Both vertical and horizontal tendons consists of 139 06 mm wires . The tension load is just below 5 MN.Each horizontal tendon stretches half a turn round the containment. The total number of horizontal tendonsis 245 and vertical tendons 153 .

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1.4 Preparation work

In order to prepare the staff that were going to do the civil some practice was done on the wall block fromRinghals 2 . The containment wall in Ringhals 2 has the same construction as it has in Ringhals 3 . On thewall block the staff trained seem drilling and other work procedures in the same manner as it was plannedto be done at Ringhals 3 .

2 .

The transfer opening

2.1 Protective wall inside containment

The work inside the containment started as soon as the plant was shut down .

After erecting scaffolding and cleaning the inside of the containment wall and the floor next to the transfe ropening, a protective wall was installed . The floor between the protective wall and the containment wal lwas covered with steel plates . The protective wall had a steel structure and covered with steel panels . Therealso was a ceiling of steel panels built between the protective wall and the containment wall .

In the protective wall there was a MEGA-door installed through which the steam generators weretransported.

2.2 Detensioning and removal of the tendons

Within the opening there were nine vertical tendons and 20 horizontal tendons which were detensioned an ddismounted. Another 56 horizontal tendons were detensioned . Theese tendons are situated below, abov eand on the other side of the containment than the opening . The work with the tendons started as soon as th eplant was shut down .

The detensioning work started with erecting of scaffolding outside the containment at the pilasters . Alsofor preparation the jacks and x/y-writers were calibrated .

For the tendons the grease cans were dismounted and the anchor parts were cleaned from grease . The jackwas coupled to tendon top end for the vertical tendons and at both ends for the horizontal . The presenttension load was checked. Then the tendon was tensioned to a load of 5 MN and the extension recorded.The detensioning of the tendon was done in two steps and the contraction recorded . The jack was removed.

Then the tendons were dismounted through special procedures, one for the vertical tendon and one for thehorizontal . The tendons were cleaned from unnecessary grease, inspected and winded up on a hydrauli cwinch .

The tendons and grease cans were stored indoors during steam generator replacement .

2.3

Cutting of the opening

The cutting of the opening 8 m x 6.6 m started as soon as the tendons in one side of the opening wa sremoved . The first operation was to drill holes with diameter 0 300 c 200 from both outside and inside a tthe same time . Totally six drilling units worked parallel . The holes were drilled almost into the steel liner.The concrete near the steel liner was chipped away . In the bottom of the opening a hydraulic jack wer emounted and special holes were made in the corners for the sliding beams which should be used fo r

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removal of the wall block. Some drilling for rebars and bolts were done for the supports for the slidingbeams and to fix the inner concrete plug to the outer .

When the drilling was done and the fuel elements were removed from containment the steel liner was cut .In the bottom corners of the block holes were cut in order to make it possible to place the sliding beams . Asleigh was mounted on the sliding beams and the wall block was lowered down on the sleigh . And beforethe last pieces of the liner was cut the wall block was secured to the sleigh .

The block which weighed about 125 tons was drawn out from the containment wall, then lifted down to th eground level and transported to a storage area .

2.4 Preparing surfaces

After removal of the wall block the work with preparing the concrete surfaces took place . Shear recesse swere cut in the concrete surfaces in order to transfer shear forces in the joint . In the surface at the top of theopening there were ventilation channels cut out to make it possible to fill up the form with concret eproperly .

The tendon ducts were cut at the level of the concrete surface and the new tendon ducts were joint wit hinserts .

The concrete around the test channels for the steel liner was cut out .

To replace the rebars that were cut, new rebars were drilled in the concrete near the outside respectively th einside . A new layer of rebars was installed between the vertical and the horizontal tendons . The drilledholes were up to 2500 mm deep . Rebars with joint nuts were grouted in the drilled holes according to aspecial procedure .

This work were done during the period when the transportation of the steam generators went on throug hthe opening .

2.5

Restoring of the steel plat e

The work with preparing the edge of the steel liner went on during the period of transportation . A newprefabricated steel liner was placed in the opening. The new steel liner was connected to the old steel linerwith two small steel plates one on the inside and one on the outside . The channel between the small stee lplates was connected to the test channels for the old steel liner. The test channels that were cut off, whendrilling the opening, were restored . Attachments for the tendon ducts had been welded to the steel liner atthe prefabrication .

2.6 Restoring the inner part of the containment wal l

The first work sequence to be done was to mount all the connecting rebars . Then the vertical and thehorizontal rebars up to 1,5 m were mounted . The first part of the formwork was mounted and the first par tof the casting could start up to 10 cm below the edge of the formwork . The working procedure wa srepeated with mounting the horizontal rebar, mounting next part of the formwork and casting for each 8 0cm.

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In the last section there were some special arrangements done to secure that the opening was properly fille dwith concrete. A pump was connected to the formwork and the concrete was pumped into the formworkuntil it came out in the ventilation opening at the top of the opening . The formwork was made of stee lplates and left after the work was finished .

2.7

Restoring the outer part of the containment wall

In order to secure a low temperature in the concrete during the hardening pipes for cooling water wer emounted at three depths in the concrete . The first row of cooling pipes was mounted close to the steel liner ,the second at approximately the same depth as the vertical tendons and the third just outside the horizonta ltendons .

The work procedure had the following parts :

Mounting the first row of cooling pipes .

Mounting vertical tendons ducts to the attachments and connecting with the existing ends of th etendons ducts .

Connecting and mounting the first layer of vertical rebars .

Connecting and mounting the first layer of horizontal rebars .

Mounting the second row of cooling pipes .

Mounting horizontal tendons ducts to the attachments and connecting with the existing ends of th etendons ducts .

Mounting the third row of cooling pipes .

Connecting all the vertical rebars and horizontal up to 1,5 m in the second layer .

Mounting all the vertical rebars and the horizontal up to 1,5 m .

The first section of the formwork was erected .

After connecting the cooling system to the cooling pipes the casting took place up to 10 cm below th etop of the formwork .

The work procedure with connecting and mounting horizontal rebars, erecting formwork and castingwas repeated each 80 cm .

To the last section of the formwork a pump was connected and the concrete was pumped into th eformwork until it came out in the ventilation opening at the top of the opening .

The concrete quality used was K50 with the temperature 5 °C. In order to prevent cracks from hightemperature during the hardening the concrete was cooled . The incoming temperature of the cooling wate rwas to 2 °C. The temperature of the cooling water coming out of the pipes was registered and compare dwith calculated values . The supervision of the temperature went on for some time after that the casting wa sready and the temperature had decreased to a normal level .

2.8 Tensioning of the tendons

The work with the mounting and tensioning of the tendons started about 12 hours after the casting wa sfinished. The first moment was to insert the tendons that are crossing the opening in the ducts . When thetendons had got through the duct the anchors were thread on each end and a button head was made at eac hwire.

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Before the tensioning work started the jacks and x/y-writers were calibrated . When the compressivestrength in the concrete in the opening had reached 36 MPa the tensioning of the tendons outside th eopening could start . To start with the tensioning of the tendons crossing the opening the compressiv estrength in the concrete must have reached 40 MPa .

The tensioning was done according to a specified procedure to the specified load and the same shims wer eplaced at both ends as before detensioning. Both the load and the elongation was recorded . When all thetendon were tensioned the grease cans were mounted the cans and tendon ducts were injected with grease .

One incident happened when the horizontal tendons were inserted in the tendon ducts . One of the duct swas damaged so that the plate in the duct was pressed to a stop in the duct . It wasn’t possible to insert thetendon. The place, where the stop occurred, was located . It was situated outside the opening. It wasnecessary to cut the concrete away to uncover the stop. It was done with water jetting . The stop was foundand the duct repaired and the concrete restored. The tendon was then inserted and tensioned .

2.9 Removing of temporary construction s

A couple of days after the casting of the concrete the removal of the temporary wall on the inside begun .The area was cleaned and the material was transported out of the containment .

When the tensioning of the tendons was finished the scaffolding was removed .

3 .

Inspection program

3.1 Contractors control program

In the scope that the contractor had there was prescribed a control program for the civil work that include dnormal control on concrete works, the drilling, cutting, welding and prefabricating the steel liner and th etendon works . A special control plan was drawn up for this purpose .

3.2 Status of the containment wall

At some incidents that had happened at other plants before, questions about the status of the containmen twall had been raised . In order to answer some of these questions a control program was established . Themain parts in the program were :

Visual inspection of all uncovered concrete surfaces to look for cracks, cavities from the casting an dother abnormalities .

• Visual inspection of uncovered rebars to look for corrosion and how the concrete has enclosed therebars .

• Visual inspection of the steel liner to look for corrosion .Take out concrete cores to examine the compressive strength, the carbonation deep and the deep of th echloride penetration.

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3.3

Results

One observation made, was a crack between the steel liner and the concrete at the inside of the steel liner .This phenomenon is known before from Ringhals 2 at the steam generator replacement . The crack widthwas approximately 1 – 2 mm and according to a special investigation the crack will remain in the area nextto the opening but disappear for the rest of the containment wall at the tensioning of the tendons .

There also was one small crack in the outer wall at the top of the opening . The crack width was less than 1mm and was probably caused by shrinkage of the concrete . The crack closed probably at the tensioning o fthe tendons . No other signs of degradation of the concrete were discovered .

There were no signs of corrosion on the rebars and the concrete had enclosed the rebars quite good .

The uncovered parts of the steel liner showed no sign of corrosion . The steel liner surface had a thin layerof cement mortar .

The compressive strength was 91 MPa as an average value . The origin compressive strength was set to 5 0MPa and samples taken by the casting showed values of 60 - 70 MPa after 91 days. There is an increase ofthe compressive strength with increasing age of the construction .

A normal value of the carbonation was 8 – 10 mm . It is about 20 years since the containment wall waspoured. Our estimate is that when the construction reach 100 years of age the carbonation have reached adepth of about 20 mm in the wall .

The result from the test of chloride gave that the concentration of chlorides is very low . The highest valuewas 0,2% chloride of the cement weight at the surface and less than 0,1% in depths of 3 – 6 cm .

The conclusion of the inspections and tests mentioned above are that the containment wall is in goo dcondition and there are nothing that indicates a degradation of the containment .

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124


In Service Inspection Programme and Long-term Monitoring o fTemelin NPP Containment Structures

Jan Malÿ, Jan gtepân - Energoprojekt Praha a .s .Czech Republic, Prague

Abstract

The paper describes monitoring systems of Temelin containment concrete structures and pre-stressingsystems. Reliable information on the actual state of containment structure is necessary for conditio nassessment as well as for detection of local defects and effects of ageing . In service inspection programmeas a main tool of preventing structural defects and damages will be discussed. Results of Temelincontainment inspections and tests will be presented .

Containment structure

The Temelin NPP is formed by two VVER 1000 MW units . At present the start-up of the 1st unit at 100%power output is underway, for the 2nd unit fuel has been loaded and the physical start-up process i sunderway. The units of the VVER 1000 MW type have a PWR (pressurised water reactor) reactor and acontainment of pre-stressed reinforced concrete . Typical cross section of the reactor building is shown i nFig. 1 .

The containment consists of a cylindrical and a dome part . Connection between the cylindrical and dom eparts is made with the help of a rigid ring beam in which the anchoring blocks of pre-stressing cables ar eplaced. The wall thickness of cylindrical shell is 1 .2m, the dome wall thickness is 1 .1m. The containmentstructure is placed on the reinforced concrete slab at a level of +13 .20m, thickness being 2 .4m. This slabcontains also supporting blocks of pre-stressing cables of the cylinder which are built-in there . Thecontainment is made of concrete, grade B40 according to the Czech standards (CSN) . The tightness of thecontainment is ensured by the steel liner of a thickness of 8 mm made of carbon steel .

A chart of the pre-stressing method is shown in Fig . 2 . The cylindrical part of the containment is pre-stressed by 96 cables running in helical direction . The cable anchors are installed in the upper part of th ering beam, the bending of the cables takes place in the slab at a level of +13 .20m. The dome part of thecontainment is pre-stressed by an orthogonal grid plan of pre-stressing cables formed by 36 cables . Alway stwo cables are conducted against each other, anchors of one cable and bending of the other one are situate don one side. The anchoring blocks are installed from the ring beam side . The cables of the cylinder anddome parts are of the same structure and cross section. Cable preservation was made with grease durin gproduction, preservation of anchors was made after pre-stressing . The anchors are protected from climaticeffects by means of the sheet covers installed .

The pre-stressing unbonded cables are conducted in polyethylene tubes . Every cable is formed by 45 0wires featuring a diameter of 5 mm. Low-relaxation wire was used for production, its yield point being1620 MPa. The initial pre-stressing force according to the design is 10 MN . On the basis of experienc eacquired during the pre-stressing of the 1st unit the anchor details were modified . These modifications haveensured better arrangement of wires on the anchor, and thus also a more even distribution of the pre-stressing force into the individual wires .

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The containment function was verified during the structure integrity test (SIT) . The test carried out on boththe 1st and 2nd units was implemented as a combined test . The function was tested for strength at a noverpressure of 460 kPa, the function was tested for tightness at an overpressure of 400 kPa . The SIT ofthe 1st unit was carried out in 1998, the SIT of the 2 nd unit was carried out in 2000 .

Overview of inspection activities during operatio n

For the purpose of ensuring full performance of the containment for the entire period of operation of th eunit there was created an inspection programme of the containment structure . The inspection of thecontainment structure consists of the assessment of measurement of sensors of the permanently installe dmeasurement systems and of the inspection of the conditions of concrete, liner and pre-stressing system .According to the frequency of the work carried out the inspections divide the work into two phases . Duringthe first phase (the first 4 years of operation) the activities are carried out in full extent . Within theframework of the following phase a part of activities is carried out on an annual basis, and another par tonce in four years .

The inspection of the containment concrete consists of the following activities :

inspection of the containment surface – to be carried out twice a year . Focused on checking fordamage, corrosion of the reinforcement system and crack development .

non-destructive concrete strength tests – carried out once a year, in the second phase once in fou ryears .

Liner checks are carried out always when it is possible to enter into the containment . The inspectionconsists of the following activities :

inspection of the coating for integrity and of the liner for damage .

non-destructive measurement of liner thickness .

check for tightness – carried out within the periodical test framework .

The checks of the pre-stressing system is carried out in the first phase of the inspection work once a year ,in the second phase once in four years . The inspection consists of the following activities :

inspection for humidity at the place of anchors and bends .

inspection for integration of preservation at the place of anchors and bends and change in chemicalproperties of grease .

Inspection of wires and anchors for damage .

checks of the pre-stressing force by lift up tests .

The above listed activities are completed with the inspection of the cable removed . The dismantling of thecable will be made three times during operation, there will always be removed two cables of the cylinderand one cable of the dome. The inspection is focused on the condition of preservation, level of corrosionand damage to individual wires and mechanical properties of wires are verified on selected samples .

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A new methodology has been developed for the lift up test . This methodology makes it possible to specifyexactly the pre-stressing forces in the anchor . during the test the pressure in the press is recorded, as well a sthe force measured by Hottinger sensors and anchor lift from the supporting block (measured by thedisplacement sensors temporarily installed) . This monitoring enables exact specification of the moment o fanchor relief and thereby also the exact determination of the pre-stressing force in the cable .

Assessment measuremen t

In order to enable the inspection of the level of the containment pre-stressing, measurement systems areinstalled permanently on the structure, and these systems measure structure deformations and pre-stressingforce in the cables . The measurement is carried out once a month, once a year the setting of the reacto rbuilding is measured . Inspection of the values measured is carried out at each measurement, a complet eassessment is made once a year .

The following measurement systems are installed on the containment :

NDS and SDM systems – these two systems consist of vibrating wire fitted during concret epouring into the containment walls . The sensors are of four types and measure concretedeformation, temperature and horizontal shift in the middle of the height of the cylindrical part o fthe containment . The containment includes more than 240 sensors which are installed in it (246 onthe 1st unit and 256 on the 2nd block) .

Hottinger system – this system is formed by strain-wire gauges stuck on the anchors of all cable sof the cylinder and of the dome, i .e . 264 anchors measured. The sensors measure force in th eanchor of the pre-stressing cables .

MEM system – the system is formed by the sensors installed on the conduits of the pre-stressin gcables . These sensors measure force in the cables by means of the magneto-elastic method . Thesensors are placed on two cables of the cylinder and of the dome . The sensors on the dome cable sare placed under the anchor and the cable bend . On cylinder cables they are placed under th eanchor and at the middle of the cylinder height, on the 1st unit the sensors are installed in the lowerpart of the cylinder as well .

HYNI system – the system measures the settlement of the reactor building by means of hydrostati clevel control . Measurement is carried out with regard to the criterion of the reactor inclination .Since the reactor building is founded on a rocky bedding, the settlement was at a minimum leve land there are virtually no changes anymore .

The distribution of sensors on the containment structure is illustrated in Fig . 3 . At present it is the concretecreep that has the largest impact on the change in the state of stress and containment deformation . Withregard to the age of the structure, the effects of the pre-stressing cable creep and of concrete shrinkage isminimum .

As an example of the values measured by vibrating wire of the NDS system, Fig . 4 illustrates thedevelopment of proportional deformations in the central part of the dome of the containment of the 2 nd unit.From the graph it is possible to see the deformation in course of pre-stressing, and subsequentl ydeformation due to concrete creep . The values measured also reflect the effects on the values measured inthe case changes in outdoor temperatures during the year . If measurement is made in a period of severalhours, the effects of outdoor temperature oscillation during the day has similar features . For the purpose o fcomparison, Fig . 5 shows the time history of temperatures in the dome during the same period.

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The example of the pre-stressing force course on the cable is shown in Fig . 6 . The graph states the value sof pre-stressing force measured with the MEM system sensors on the cylinder of the 1st unit, cable no . 21a.From the graph it is possible to see the decrease in the pre-stressing force along the cable length due tofriction, as well as decrease in the pre-stressing force over time as a result of concrete creep .

The comparison of measurement Hottinger, MEM and lift up tests is in Fig . 7 and 8 . The graphs providecourses of regression of the average value of the pre-stressing force specified by the systems Hottinger andMEM (for sensors under the cable anchor) on the cylinder and on the dome of the 2 nd unit. The graphs alsostate average values of pre-stressing force determined by the lift up test after 2000 hours from the pre-stressing and before the SIT .

Conclusion

The methodology of the in-service containment inspection presented in this paper enables us to inspect an dmonitor the containment of the 1st and 2nd units of the Temelin NPP, and provides a guarantee that thecontainment structure is able to perform the function of the last safety barrier .

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Fig. 1 Cross section of the reactor buildin g

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Fig. 2 Scheme of pre-stressing of containment

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5 ensors of system NDS and 5 DM

5 ensors of system Hottinger and ME M

in the wall of containment

-

on the cables of containmen t

Fig. 3 Distribution of sensors on the containment structure

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y

~

I

N N Co Co CD CD N

Fig. 4 Proportional deformations in the central part of the dome of the containment of the 2nd unit measuredby string strain gauges of the NDS system, sensor type PSAS .

Fig. 5 The time history of temperatures in the dome during the same period as in Fig . 4 .

Average of int . sensors Average of ext . sensors

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Fig. 6 The example of the pre-stressing force measured with the MEM system sensors - the cylinder of th e1 st unit, cable no . 21 a

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1 0

9 . 5

z

d 9vôLL

8 .5

8

01 .07.1996 01 .07.1997 01 .07 .1998 01 .07 .19997LPH

01 .07.2000 01 .07.2001 01 .07 .200 2

Regresion of Hottinger cyl . -Regresion of MEM cyl . --O--Lift-up cyl .

Fig. 7 The comparison of measurement Hottinger, MEM and lift up tests - the cylinder of the 2nd uni t

Regresion of Hottiger dome -Regres ion of ME M donne t Lift-ap dom e

Fig. 8 The comparison of measurement Hottinger, MEM and lift up tests - the dome of the 2nd unit

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REPAIR CRITERIA & METHODS OF REPAIR FOR CONCRETE STRUCTURES O FNUCLEAR POWER PLANTS

PARTICULAR APPLICATION ON NATURAL DRAUGHT COOLING TOWERS IN BELGIUM

R. Lasudry, Principal Enginee rBELGATOMGTEC DepartmentCivil Works BranchAvenue Ariane 7, B1200 Brussel sTel : 32 2 773 81 3 4Fax : 32 2 773 89 7 0e-mail : roland.lasudry@tractebel .be

Abstract

A previous paper was presented at the OCDE Workshop held 22 – 23 March 2000 in Brussels explainingdifferent aspects of the techniques used for “Instrumentation and monitoring of natural draught coolin gtowers in Belgium” .

These monitoring and preventive techniques are now applied in Belgium since already more then 10 year sby Tractebel on the towers of the nuclear plants .

These huge constructions have to sustain considerable physical, chemical and biological loads . As one canfigure out, and as years go by, these inspections showed deterioration of which type, progress, quantit yeventually led to the need of repairing these structures .

The present paper goes over 4 main different sorts of defects (beam supports breaking, fast carbonatio nrate, concrete porosity, and a series of local deteriorations like insufficient concrete cover, cracking, gravelpockets, corroded reinforcement) encountered on 3 cooling towers situated in Belgium, and affecting theshell as well as the inner structures .

The diagnosis, the choice of the appropriate repair techniques and products which will avoid having to fac emuch higher costs in the future are explained .

It also gives an illustration of the works carried on site and points out the uncommon and complex aspect sthe treatment of such a construction implies (planning, both horizontal and vertical curved shape, works a tgreat height, …) .

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1.

Introduction

A cooling tower, whether at a fossil-fired or a nuclear power plant, where the circulation water is coole dafter having left the condenser, is a main component as it provides the cold source in the thermodynami ccycle of the turbine .

On account of the water flow rate to be treated and the volume of air required, the cooling towers, built ofreinforced concrete, are structures of hyperboloid revolution that are very impressive by their size an dshape (up to 160 m high for the large installations) . The developed surface area of the thin concrete shell ofthese towers may reach several tens of thousands of square metres . By their nature, the cooling towers haveto sustain considerable physical, chemical and biological loads .

The inner structure is a construction of beams and columns designed in order to provide the support of theair/water exchange material allowing this construction to perform its cooling function .

While for a long time concrete constructions were considered capable of defying the years withou tproblems, this simplistic approach has since been abandoned, as it is now recognised that, like othe rmaterials, concrete is affected by ageing and by various illnesses that need to be treated and kept undercontrol if the concrete is to reach its optimal operational life . These constructions were generally orderedby the operators as a turnkey component and, as such, were not verified (calculations and works) by th eengineering consultants as Tractebel .

Moreover, in the seventies, several cooling towers even collapsed, like in Ferrybridge (England) o rBouchain (France) reminding all structural and electrical engineers that a major failure of this essentialpower plant component was always possible .

2.

Technical approach

2.1

General description and diagnosi s

The concrete of a cooling tower, exposed to the aggressions of its environment and to the operatin gconditions, presents also a number of characteristics that are specific to it as compared to other structures.Therefore, investigations are first required to identify the causes of the deteriorations, as the causes wil ldictate the methods and materials required for a successful restoration .

The origins of the deteriorations may be chemical, physical or biological : wind action, the structure’s owngravity, inside/outside temperature gradient, insulation differential, rain action, air pollution, soil-structur einteraction, water vapour and proliferation of algae or mosses on the shell leading to cracking, deficiencie sinherent to the concrete such as the presence of gravel nests, chipping and spalling, deteriorations causedby carbonation, alkali-silica reaction or by chlorides .

In the case of a cooling tower, all these deteriorations make fragile a structure which is constantl ysurrounded by an atmosphere saturated with water vapour and which is exposed to thermal differentia lstresses the major part of its existence .

These environmental conditions lead to the slow deteriorations of the tower’s components : reinforcementcorrosion, concrete leaching .

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2.2

Treatment guidelines

Accordingly, treating a component such as the shell of a cooling tower requires many investigations, cor esampling at significant locations, laboratory analyses, selection and validation of particular treatmen tprocesses and procedures that are suitable, considering also the access constraints and the size of th esurfaces needing treatment. In the case of serious degradation of the concrete performances, a ne wcalculation of the shell might need to be performed that could lead to strengthening measures bein gprescribed .

In general and if the chemical deteriorations are limited, treatment involves repair of the physica ldeteriorations, followed by preventive treatment of the concrete against aggressive elements .

Repair of the physical deteriorations comprises the treatment of the reinforcement and the replacement o fthe damaged concrete so as to return the tower as close as possible to its initial condition .

This first action is then followed by surface treatment . Taking into account the geometrical characteristic sand the deformation of the shell, this treatment usually involves applying to the inner face, which is the on eexposed to the water vapour, a treatment more impervious than the one applied to the outer face . As aresult, the inner face is sealed off against ingress of water into the concrete, and the concrete is allowed to"breathe" on the outside. Also, as the two faces are treated, penetration of carbon gases and othe raggressive elements is prevented and chemical reactions can be stopped .

In practice the review of the way the treatment is to be implemented is completed by a comparativ eappraisal of the methods of accessing any point of the inside and outside of the shell . Complex problem sare taken into consideration in this review, relating to the height of the structure, its both horizontal an dvertical curved shape and the necessity of supplying to any point the liquids and products under require dpressure for the treatment . For instance, high pressure water-jet cleaning requires water at 400 bar (or evenmore) to be provided at great height. Also, all the means of access have to present sufficient safetyregarding operators and products, while making possible the rapid progress and quality of the work .

Finally, as the works are only performed during a stop of the tower, usually meaning lowering or stoppingthe production, these equipments are to be suited (in number, speed and ease of use) with the shortes tpossible working schedule .

2.3

Encountered problems

The 3 first examples concern degradation affecting the shell of 3 power plant cooling towers (Ruien 5/6 ,Doel 3 and Tihange 3) .

The last example concerns problems affecting the inner structure of the Tihange 3 cooling tower .

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3.

Ruien power plant units 5/6

3.1

Main features

Cross-flow type cooling tower built in the years 1972-1973 :- diameter at the ground level ( cold water basin not included) : 60.90 m- diameter at the lintel : 54.30 m- diameter at the throat : 41 .60 m- diameter at the top : 48 .50 m- height at the lintel : 16 .15 m- height at the throat : 62 .58 m- total height : 94 .65 m- shell thickness at the lintel (varying up to 28 m) : 0 .70 m- shell thickness from elev .28 m to the top : 0 .12 mThe reinforcement is made of only one centred reinforcement layer .

As significant vertical cracks appeared after a few years, the builder placed at his own costs under the 10 -year guarantee, an additional external reinforcement to avoid a potential collapse of the structure . Thisinvolved placing 42 cables T13 (each tensioned with 30 kN) every 0 .80 m from the throat to the top .

3.2

Problem description - diagnosis

The cooling tower of the Ruien power plant was found to be in the following condition :• shell : cracking (partly passing through), local deteriorations, gravel nests (some going right throug h

the shell), corroded reinforcement, uneven concrete surface, though fairly good concrete quality inthe sound parts of the shell, carbonation depth < 15 mm (pict . 1) .

shell supports : cracking and deteriorations under chemical attack .

The nature, the quantity and the size of the shell cracks were such that the initially monolithic structure ha dnow to be considered as an articulated one .Also, the concrete needed treatment to stop leaching and protect the reinforcement .

3.3 .

Treatment definitio n

In a first phase, core samples were taken from the shell in order to identify the physical and chemica lproperties of the concrete, so that intended treatments could be assessed and their chances of succes sevaluated.

After this, the phases of the treatment were defined :

complete cleaning of the inner and outer surfaces (pressure jetting with water and/or sandblasting)(pict. 2) ;

removal of all the portions of poor quality concrete (in and out) ;

treatment of reinforcement (in and out) : cleaning (all around the bar), replacement when necessaryand protection against corrosion with an hydraulic-based product ;

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replacement of the removed concrete with a new hydraulic cement-based compound, filling of th egrid nests and correcting of the unevenness with the same product (in and out) ;

complete shaping (inside only) with a fine polymer improved hydraulic rendering cement (6 kg/m2) ;

inner face protection : application of impervious coats that would remain flexible enough to follo wstructural deformation . On an impregnation epoxy resin coat, a polyurethane-based multi-laye rflexible system (2 mm) was laid, protected afterwards with a final coat (40 µm) providing the UVradiation protection ;

outer face protection: application of protective coats witch allow the concrete to breathe and whic hare not sensitive to structural deformation . A hydraulic elastic polymer mortar was applied in tw ophases of 1,5 kg/m2 and 3 kg/m2, the first layer applied with a trowel, the second gun-applied .

These coats provide the concrete with a barrier against attack by various substances . Moreover, theapplication of cement-based products brings a complementary quantity of "fresh" lime that contributes t ostabilize and partly restore the alkaline properties of the existing concrete, improving thus the protection o fthe reinforcement and restoring the intrinsic strength of the concrete .

In a second phase, tests were performed on selected areas in order to determine the corrosion level of thereinforcement and validate the product application methods that proved the most effective .

3.4.

Works

A particular problem was faced with this cooling tower due to the pre-cast cables being located outside th eupper half of the concrete shell . This had to be taken into account when carrying out the various operation s(movement of hung scaffolding, pressure cleaning, application of the coating products) (pict . 3).

Concerning the shell supports, the bases of the columns were injected with a sizing epoxy-based resin .However the small opening of the cracks of the columns did not make their injection possible . Thecolumns have been treated with the same product as the inner face, however in two layers only .

The works have been carried out under cover of a quality assurance program and controlled by the expert sof an insurance company allowing it to be covered by a 10-year full warranty conditional to regula rinspections .

Representing the treatment of 12,000 m2 per face plus the supporting columns, they have been carried outwithin a 3 months period (of which only one month of real working days due to bad weather conditions )for the inner face and about one year for the outer face (with an average working ratio of 45%, winter sto pincluded) .

4.

DOEL 3 Power plant

4.1 .

Main features

Counter-flow type cooling tower build in the years 1981-1983 :- diameter at the ground level ( cold water basin not included)

: 141 .70 m- diameter at the lintel

: 133 .88 m- diameter at the throat

: 76 .64 m- diameter at the top

: 83 .56 m

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- height at the lintel : 12 .70 m- height at the throat : 108 .70 m- total height : 167 .28 m- shell thickness at the lintel (varying along the first 10 m) : 0 .85 to 0 .25 m- shell thickness from elev . 10 m to 104 m : 0 .25 to 0 .18 m- shell thickness from elev . 104 m to the top : 0 .18 m

The reinforcement is made of two reinforcement layers, the theoretical concrete cover is larger than 40mm .

4.2 .

Problem description - diagnosis

The cooling tower of Doel 3 is affected by an illness that results of its shell outer face incurring a too fas tcarbonation rate (penetration of carbon dioxide gas in the concrete). This phenomenon, which wa sdemonstrated by a series of tests carried out in Belgian laboratories, leads to a modification of th ealkalinity of the concrete that protects the reinforcement . With time the reinforcements get corroded an dcause serious deteriorations as the concrete bursts . In our case, the penetration depth was about 30 to 3 5mm, which is almost equal the concrete cover of the reinforcement .

4.3 .

Treatment definitio n

The propagation of carbonation can be arrested by the application of a coat that prevents the carbo ndioxide gas penetrating into the concrete .The suitable products are selected following in situ tests and manufacturer’s specifications . The treatmentof the outer face of the wall is however not sufficient . Further investigations resulted in an imperviou scoating being planned for the inner surface so as to avoid ingress into and diffusion within the concrete ofwater vapour, since this would have induced flaking of the paint on the outer side . As the concrete was notmuch cracked, there was no need for flexible coating . An impervious epoxy-based system was chosen forthe inner protection, it is applied in 2 layers (non pigmented impregnation layer of about 500 g/m2 an dpigmented final coat of about 300 g/m2) .

The outer protection consists of 3 layers of a one component synthetic (plastic mixture based on PV Ccopolymers) paint (300, 400 and 300 g/m2 of different shades) .

The latter effectively protects the concrete by preventing carbon dioxide penetration (CO2 diffusion

coefficient µ = 3 .05 106) whilst allowing the concrete to breathe (permeable to air and water vapour) an doffering a satisfactory resistance against ageing (chiefly UV radiation) . Because of the unpredictabl eweather in our countries, the paint was chosen (solvent-based) so that it could be applied on wet surface sand drying time would be short (1 to 2 hours) .

4.4.

Works

Most of the treatment operations are similar to those already described, though a number of particularitie sexist at Doel 3 .

For instance, as the outer face of the shell had its concrete roughened in alternance by two types of verticalribs (the one long and slender and the other short and broad), such surface offering better behaviour underwind loads. After a few years, the corrosion of the longitudinal reinforcement that had been placed in th eslender ribs (which had a very thin (5 to 15 mm) concrete cover) induced a general cracking of these rib s(pict . 5) . A new calculation of the shell showed that the slender ribs could be completely removed from th etower. This was in fact less expensive and more feasible than trying to repair them . These works were

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carried out 2 years before the treatment of the complete shell was considered . A new problem occurredbecause of the remains of the reinforcement that previously linked the ribs to the shell . This steel had beencut close to the shell and was corroded due to an insufficient protection . These countless rust point srequired a particular treatment method that was defined after detailed trials and tests (pict . 6) .

Another particularity resides in the fact that the plant of Doel (4 units) is only provided with 2 coolin gtowers, each one serving 2 units (D1 coupled with D4 and D2 with D3) . The stop of two units is thu srequired if work is to be performed inside a tower unless the thermal conditions of the outfall allows thecirculation water of one unit to bypass the tower while the other is stopped . In our case, the entire treatmentinside the cooling tower had to be carried out within the strictly limited period that unit 3 was shut dow nfor steam generator replacement. In this short period had to be covered the assembly/disassembly of thefixed and mobile scaffolding installations, the automatic remote-controlled and manual cleaning devices ,and of course all the cleaning, repair and painting operations . Furthermore, the presence of 2 stiffeningrings on the inner face of the tower (each protruding by some 1 .25 m) did not make things easier (pict. 4) .With some precautions, the painting of the inside was done using an airless spray gun . Thanks to goodforward planning, the operation was completed successfully within the six weeks time allocated .

On the contrary, the restoration work on the outer face of the shell took much longer then expected, as thenecessity appeared for corrective treatment of the ribs . Dealing also with the weather and temperatur econditions (leading to an average working ratio of 45%, winter stop included), the work could not b ecompleted in less than one and a half years despite the simultaneous use of 4 cradles . In order to obtain an

homogeneous final shade, the last coat (representing 48,000 m2 and manufactured in one batch) wa snevertheless applied in only one week .

The whole work was also controlled by the experts of an insurance company allowing it to be covered by a10-year full warranty conditional to regular inspections . The first of these inspections took place after oneyear and did not reveal any visible deterioration .

5.

Tihange 3 Power plant

5.1 .

Main features

Cross-flow type cooling tower built in the years 1981 - 1983 :- diameter at the ground level (cold water basin not included)

: 120,5 m- diameter at the lintel

: 111,30 m- diameter at the throat

: 62,9 m- diameter at the crown

: 67,8 m- height at the lintel

: 8,7 m- height at the throat

: 109,3 m- total height

: 157,5 m- shell thickness at the lintel (varying along the first 10 m)

: 0,86 to 0,21 m- shell thickness from level 10 m to 100 m

: 0,21 m- shell thickness from level 100 m to 115 m

: 0,21 to 0,17 m- shell thickness from level 115 m to 140 m

: 0,17 m- shell thickness from level 140 m to the top

: 0,17 to 0,50 m

The steel reinforcement is made of two layers, the theoretical concrete cover is larger than or equal to 40mm .

14 1


5.2 .

Damage description - diagnosis

Since a few years, the concrete of the external side of the shell of Tihange 3 cooling tower became ochre-coloured. The outer side of the shell had become ochre several years already below the throat, due to thi spart of the face being exposed to rainfall . More recently this was increased by percolation which formed apermanent wet patch presenting traces of calcite, several hundreds of m² covering the first 15 casting ring s(each of 1 .5 m high) .

In addition to the inspection routinely performed in the scope of the monitoring program, several particularanalysis were carried out to focus on the phenomenon which induced this particular pathology . Theseconsisted of :

a chemical, petrography and microscopic analysis of core samples taken within and outside the we tpatches ;

a chemical analysis (particular for Fe content) and microscopic examination of algae fragments take nfrom the internal face of the shell, and for comparison purposes from another cooling tower at the sam esite;

a corrosion study of the reinforcement by electrical potential measurements in the wet patches and in atotally opposite area;

The overall analysis of these two sets of investigations permitted drawing coherent conclusions which canbe outlined as follows :

the absence of correlation between the deformation and the cracking hints at the concrete itself being a tfault;

the reinforcement corrosion was found low except in the wet patches where it can be said medium tostrong ;

the concrete porosity is very high, its density and compressive strength are low ;

the ochre colouring does not reveal reinforcement corrosion, but appears to result from a mineralogica ltransformation of ferrous particles present in the cement, resulting of a drop in pH due to concretecarbonation . The carbonation front matches exactly the coloured zone .

The percolation observed in the wet area in fact results from the too high porosity of the concrete . Thepermeability throughout the mass of the concrete induces the percolation, the engine of this being th epressure gradient between the inside and the outside of the tower . .

5.3 .

Determination of the treatment

Water percolation jeopardizes the durability of the tower as it, in the longer time, results in reinforcemen tcorrosion and a washing away of the concrete mix constituents .

We therefore found it necessary to fight percolation by impeding the water to penetrate into the concrete ,by applying an inner waterproofing coating composed of two layers of epoxy, in addition to the primer .This treatment is adequate because the cracking is low . Several other reasons (limited time, importance o fpermanent cost, technical risks, …) resulted in the decision being taken to treat all the entire inner surfacerather than just the affected area.

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No treatment was advised for the outer face as it is important to be able to ascertain the efficiency of theinner coating . The outer face has to be left as is not to mask a possible evolution. Also no technicalarguments (since carbonation and corrosion are low) can justify outer treatment which as we know is ver ycostly, and moreover, it allows the concrete to "breathe" .

5.4.

Works

The work was quite similar to that done on the inside of the Doel 3 cooling tower shell, applying the samemethods and essentially requiring a sufficiently long period of stoppage of the plant (pict . 7) .

A steam generator replacement which was planned at Ti3 in the summer `98 offered that suitable windowto treat the cooling tower . However due to other maintenance work to the tower, only 6 weeks wer eavailable to achieve this treatment .

The challenge was met : the first site meeting was held on 10 June and the last lick of paint was applied o n20 July, despite 12 days in all being lost due to bad weather .

The attached tables 1 and 2 illustrate the progress of the repair and coating of the 37 .600 m2 of the innerface of the structure, which involved applying some 37T of product . Appendix 3 gives an idea of the poo rweather conditions (essentially rainy days, but wind can be embarrassing as wel), and their impact i sreflected on the progress curve .

Like at Ruien and Doel the work was covered by Quality Assurance requirements and supervised b yexperts so as to insure the 10 year guarantee .

In addition to the practical aspects of this type of work which are similar to the already mentione dexamples, one point was analyzed particularly : the structural stability of the crown walkway from whic hthe painter's cradles had to be suspended. Depending on the orientation and the suspension points, thes ecradles induce tilting loads of several ton/m . Indeed this tower inspection walkway is not an integral part o fthe shell . It is composed of a series of prefabricated "U" elements, each fixed individually by means of 2bolts on the top edge of the shell and 1 bent rebar at each end . These U elements are each 3 .4 m long, arestraight and are spaced 12 cm from each other, forming a polygon on top of the crown (pict . 8) .

The checking of the stability of these boxes was favourable providing their anchoring, which were alread y15 years old, had retained their original strength .

This was verified by :

• tapping and visual inspection of the seals (absence of cracking, rust, dull sound )

• core samples to test the quality of the material

verify the compliance between the reinforced drawings and the execution, through a magnetic "X Ray"of the elements in order to check the position, the number, the diameter and the cover of thereinforcement .

The confirmation of the quality of the concrete and the suitability of the reinforcement compared to thecalculation allowed to go ahead with the treatment with full knowledge of the facts .

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6.

TIHANGE 3

6.1 .

Introduction

Not only the shells of these towers can be source of problems !

As said before, the cooling tower is a heat exchanger the function of which is to cool down the water of th ethird circuit of a nuclear power plant, cooling itself the turbine condenser (circulation water) .

This is obtained by a heat exchange between air (up-draught) and water . The cooling water is sprayed onthe inner horizontal surface (representing about 10 .000 m², i .e . about 3 football fields!) by a net of channel sand pipes. It is sprinkled above an exchange material (called packing) across which it trickles (pict . 9) . Innormal conditions, the water flow rate represents an amount of water of 4 l/m²/s, in winter conditions thi svalue can be doubled.

The water is collected at the foot of the tower in a basin and then flows by gravity to the return circuit, to adischarge in the river or to the pumping station depending on the thermal conditions .

On Thursday 3 January 2002, the break of 2 brackets of the supporting structure resulted in the collapse o f2 beams and the packing they were supporting (pict . 10) . The flood entrained the debris which was stoppedby the protection screens placed at the entry of the returning pipes . The inlet of the pipes became rapidlyclogged and eventually the basin flowed over . Due to the geographical position of the tower, the overflo wflooded to some of the NPP buildings .

6.2 .

Finding s

The NPP operator ELECTRABEL asked TRACTEBEL Energy Engineering to analyse the accident . Thenext day some quick investigations were made (pict . 11) :

- the break affects two brackets embedded at the lower face of principal beams that supported th esecondary beams, themselves supporting the packing;

- the affected brackets are anchored at the lower face of the principal beams as if they were suspended ;- the reinforcement steel that became bare and the rupture surface do not show any trace of corrosion o r

dirt;- the concrete of the rupture surface feels sound ;- the two fallen beams are situated in the same area, present the same constructive characteristic s

(geometry, supports, …), provide an identical function and incurred the same type of break .

6.3 .

Analysi s

As the rupture surface was clean and there is no steel corrosion, one may consider that the failure wassudden (brittle break, like a lump of sugar) and corresponds to a shear break of the concrete . The concretecharacteristics seemed good, core samples have nevertheless been taken in the wrecked brackets to verif ythis aspect . Results are satisfying .

Other causes were thus to be searched in an inadequate match between the design and the actual loa dapplied to the bracket .

144


The consultants have investigated in the following directions while the cooling tower was stopped (in fact ,the NPP can work with the tower by-pass, the circulation water is not cooled as it is sent to the basin befor ebeing sprayed in the sprinkling system) :

- loading condition of the structure- resistance capacities of the structur e- visual inspection of the structure

a) Loading state

An eventual overload of the bracket can only result from an excess of weight applied on the supporte dbeam, originating from the packing. This overweight of the packing may result from its clogging or winte rservice conditions (to minimize freezing problems, the flow is only distributed on the periphery of thepacking, which means that for a same flow rate, the concerned wet surface receives a doubled flow ,increasing the water weight as well) .

Measurements have been carried out on samples of the packing near the wrecked area . A new problemarises because the measure is made on a dry packing as the cooling tower is stopped, and therefore th ewater quantity has to be estimated.

The packing itself is made of different elements (part of their weight doesn’t change, while other parts o fthe packing can be fouled, increasing their weight) . Considering a dead load of about 0.5 kN/m³, thesurprise was great when it was discovered that some packing elements weighed more than 2 kN/m ³ . Theproblem was then to evaluate the correspondent weight of water : part of it trickling across the packing ,part retained by the clog matter . A variation formula has been used to do this, considering that the quantityof water is a function of the degree of clogging, and that it can reach up to 1,5 times the dry weight of clo gmatter in the worst case.

b) Structural analysis

Concurrent with this analysis, a complete study of the structural elements was carried out, taking int oaccount the initial design calculation codes, to determine the available safety margin for each element onbasis of the as built drawings . This study has enabled us to point out which of the elements wer etheoretically the most at risk .

c) Visual inspection

Based on the above mentioned study, a visual inspection was performed on the “most at risk” elements .The discovery of unexpected damage called for a comprehensive visual inspection involving more than athousand spots (pict . 12 & 13) .

d) Load test of the wrecked beam s

The two fallen beams have been examined and showed, over their entire length, regular cracking at theinferior face, going up on each side to mid-height . They were taken to a laboratory where they were testedto verify if the load they had supported had changed their future behaviour (still elastic or not) .

145


6.4

Results and diagnose

6.4.1 .

Packing support structure (low level)

The failure observed is due to an overload of the packing . This overload has broken 2 brackets of aseries of 24 supporting the longest beams of the tower . Their visual inspection showed a generalizedcracking .

The tests performed on the 2 beams have shown that they were still in good shape (however they ha dfallen down!), breaking by flexion at their design load.

The visual inspection of the rest of the structure highlighted the presence of degradations (mostly a tthe extremities of the beams and/or the brackets), some of them being severe enough to represent adanger for the structure .

6.4.2 .

Supports of the spraying system (upper level )

The same kind of degradation was observed randomly distributed on 10% of the supports (pict . 14 &15). As our theoretical study revealed, the thermal effects due to the temperature variation s(stop/service conditions, day/night, sun/shadow, winter/summer, …) develop horizontal loads whos eimportance is close to the vertical loads . Originally nothing was planned to limit concrete/concret efriction between the beam and its support.

To summarize, two main causes were demonstrated :

- the clogging of the packing, leading to its overweight ;- an underestimation of the horizontal loads caused by a lack of support material .

6.5.

Treatment and repair works

Twenty-two new steel brackets were designed to support both the initial one and its beam . They have beenordered, fabricated and placed within 3 weeks on the lower level to replace the brackets wrecked byoverweight of the packing (pict . 16) .

The brackets and beams whose extremities are severely cracked have been supported by scaffolding tha twill be left in place until the next stop .

At the upper level, the configuration of the brackets allowed us to design another steel system an dreinforce beam and brackets that were cracked (pict . 17). These supports were designed, ordered,fabricated and placed within a week .

6.6 .

Conclusion s

This last example shows once more, if necessary, the need of regular inspections and/or monitoring toavoid the risks of the occurrence of severe and costly damages, possibly even human consequences .

146


In this case, we proposed to the operator to monitor the packing weight by two means : continuou sweighing of selected volumes of packing and by equipping the two new beams (replacing the wrecke dbeams) with strain gauges providing a continuous measure of the actual weight carried by the structure .The study that was made gave the opportunity to set the adequate levels of alarm .The water level of the basin shall also be monitored to avoid another flood .

7.

Appendices and pictures

7.1 .

Ruien ¾ : repair and painting works

Picture 1

147


Picture 2

7.2 .

Doel 3 : painting works

Picture 4


Picture 5

Picture 6

149


7.3 .

Tihange 3 : painting works

Appendix 1Appendix 2


Appendix 3

Picture 7

151


Picture 8

7.4.

Tihange 3 : brackets breaking

Picture 9

152


Picture 1 0

Picture 11

153


Picture 1 2

Picture 13

154


Picture 14

Picture 15

155


Picture 1 6

Picture 17

156



L.M. Smith British Energy Generation (UK) Lt d

Abstract

The examination, inspection, maintenance and testing of all plant and structures that may affect nuclea rsafety on Nuclear Power Plants is of prime importance. It is essential that all nuclear safety relatedstructures must be maintained throughout their operational life in such a way that they are always fit fo rpurpose and capable of meeting their nuclear safety role as required . In order to do this they must beexamined, inspected and tested in a manner and at a frequency which is adequate to confirm that th estructural integrity, performance and reliability claims made in the safety case continue to be me tthroughout the operational life of the station . The performance criteria for nuclear safety related structure smust be determined on the basis of the duty required of each structure at each of the stages of the lifetim eof the facility . Consequently, detailed procedures exist for the routine inspection of these structures i nnormal service .

In the event of a fire on a nuclear power plant the standard inspection procedures may no longer beapplicable and the Licensee would have to demonstrate that any nuclear safety related structures in the are aof the fire are still fit for purpose and capable of meeting their nuclear safety role .

Following a fire in an area that could potentially affect the condition or performance of a structure, aspecific procedure would be written to cover the examination of that structure, the method and criteria fo rits assessment and acceptance criteria for its continued use. The specific post-fire inspection procedur ewould be based on the normal in-service inspection procedure amended to take account of the extent of thedamaged area . All areas where structural failure could cause damage to, or failure of, nuclear safety relate dequipment would be addressed. The resulting inspection report would include lists of the defects found an dgive recommendations on any remedial works required .

This paper considers the factors that would be applicable to specific post-fire inspection procedures for th einspection and assessment of nuclear safety related structures .

157



Introduction

As fire is an identified risk on Nuclear Power Plants (NPPs), it is considered at the design stage an dpreventative measures, segregation and suppression systems are built into the plant design in order to limi tthe effects of fire should it occur . However, although this greatly reduces the likelihood and effects o fpotential fires it does not totally remove the risk of one taking place .

This paper is based on circumstances as they exist in the UK but the general principles are equallyapplicable to other locations . Should a fire occur, the post-fire assessment of structures on UK NPPs woul dbe treated as a special investigation not covered by normal standing procedures . This document onlyprovides a brief over view of the subject . For more detailed information on the effects of fire on th estructural materials involved the references given at the end of the paper should be consulted [2-10] .

General Background

Each site in the UK is covered by a Nuclear Site Licence which contains 36 standard conditions whic hmust be met by the operator and is issued under the provisions of the Nuclear Installations Act 1965 b yHM Nuclear Installations Inspectorate . The rules are not prescriptive and the licensee retains absolut eresponsibility for nuclear safety under UK law . [1 ]

The regulatory requirements have major implications for the in-service inspection regime, principall ythrough site licence condition 28, which covers the examination, inspection, maintenance and testing of al lplant and structures that may affect nuclear safety . It is essential that all nuclear safety related structure smust be maintained throughout their operational life in such a way that they are always fit for purpose an dcapable of meeting their nuclear safety role as required by a detailed safety case, which is identified in thelicence conditions . In order to do this they must be examined, inspected and tested in a manner and at afrequency which is adequate to confirm that the structural integrity, performance and reliability claim smade in the safety case continue to be met throughout the operational life of the station [2] . Theperformance criteria for nuclear safety related structures must be determined on the basis of the dut yrequired of each structure at each of the stages of the lifetime of the facility .

In normal operation, a list of nuclear safety related structures and procedures for the inspection of each o fthem are prepared from examination of the Station Safety Report . The inspections are visual in the firs tinstance and checklists are prepared for each building or area of structure from detail drawings andpreliminary inspections that give guidance to the inspection team .

Under the terms of Site Licence Condition 28, after a fire, the Licensee would have to demonstrate that an ynuclear safety related structures in the area of the fire are still fit for purpose and capable of meeting thei rnuclear safety role as required by the safety case . In the event of a fire in an area that could potentiallyaffect the condition or performance of a structure, a specific procedure would be written to cover th eexamination of that structure, the method and criteria for its assessment and acceptance criteria for it scontinued use . As in the assessment of damaged areas of conventional structures, personnel safety woul dbe of high importance and require a risk assessment to be carried out but on NPPs there is the additionalrequirement to ensure radiological safety . This will require survey and monitoring of the area and a writtensystem of work to keep any potential radiological dose to personnel to a level that is as low as reasonabl ypracticable .

158


The specific post-fire inspection procedure would be based on the normal in-service inspection procedur eamended to take account of the extent of the damaged area . All areas where structural failure could causedamage to, or failure of, nuclear safety related equipment would be addressed. The resulting inspectionreport would include lists of the defects found and give recommendations on any remedial works required .IAEA Safety Report No .8 (Preparation of fire hazard analyses for NPPs) [11]covers the need for a fir epreplan. It may be argued that there is also a need for a fire postplan prepared in advance to cover th eprocedures and methods to be used to assess the immediate safety of nuclear safety related structures (i nthe order of their importance to nuclear safety) following a fire .

On British Energy NPPs, defects are defined in three nuclear safety categories ; Category 1 : Affectingsafety - repair required immediately ; Category 2: Not affecting safety - repair required as soon as possibleto prevent further deterioration; and Category 3 : Not affecting safety - repair carried out under norma lstation maintenance programme. The categorisation is based initially on an assessment made by th einspection engineer, taking into account the current safety case and overall structural integrity . Identifieddefects would be included (with normal inspection defects) in a monitoring database which allows eas yprogress tracking and identification of faults .

In addition to the defect classification given above, a damage classification of the effects of the fire on th estructural members should be carried out to allow a structural assessment to be carried out . Tabulateddamage classifications are useful in this regard (Table 3 [9]) .

When assessing a fire damaged structure the points given below should be taken into consideration .

Fire Damage Assessment

Where possible detailed information concerning the use and occupancy of a structure should be obtainedprior to a post-fire investigation to enable an estimate of the fire loading to be determined. Station Logs ,Fire brigade reports and eyewitness accounts are useful for determining the duration and course of the fire .Before the structural investigation is carried out, as-built construction drawings should be obtaine dshowing all the main members of the structure and the form of construction . Care must be exercised toensure that the effects of all previous plant modifications and repairs have been included in the informatio ncollated [3] . Previous in-service inspection reports can provide useful information on the pre-fire conditionof the structure and normal in-service inspection procedures identify critical areas of the structure .

A variety of different methods, ranging from visual survey through to non-destructive examination an dcomprehensive instrumentation, is available to the engineer to allow condition monitoring and lifetim emanagement of structures to be implemented. Structural investigations should, where possible, utilise non -invasive methods such as visual surveys or non-destructive testing although there are limitations to th einformation that may be obtained from such sources . Where it exists, instrumentation may provide usefu linformation on the performance of materials and structures . Invasive investigations involving the remova lof a sample or specimen for testing should only be used where simpler methods cannot provide th erequired information . Careful attention must be paid to the limitation of damage caused during sampl ecollection and reinstatement of the structure must be given detailed consideration and executed correctly .

The method of investigation that is chosen in any particular post-fire situation for the assessment of fir edamage to concrete structures will obviously depend on the accuracy of the results required . Each methodhas its advantages and disadvantages but every method, if properly used with sufficient care, may give a nassessment of the fire damage to the structure. No one method is entirely free from error and in most case sa combination of tests will be used depending on the importance and cost of the reinstatement of thestructure, the time allowed for repair and whether demolition and replacement is a feasible option .

159


The examination of fire debris is a very important stage in any post-fire investigation as it allows an overal lpicture of fire severity to be gained by the investigator . This is especially useful if combined with visua ldamage classification and a sounding survey of the structure . The main drawback with debris surveys i sthat the final position of thermal indicators may bear little relationship to their position at the time of th efire if any debris clearance has taken place before examination .

Pre-fire defects must be isolated from fire damage and noted specifically as such along with othe rinfluences such as blast or explosion damage (see [3, 9]) .

Photographs are a most important part of any post-fire survey as they provide a visual record and ma yyield useful information at later stages during investigation or reinstatement should problems b eencountered .

Concrete Structures

Colour changes in concrete and aggregates may be used as a valuable guide in fire damage assessment i ncases where significant changes occur due to thermal exposure . In some instances, however, a lack ofcolour change may not be taken as inferring that concrete is unaffected by high fire temperatures and careshould be exercised especially in cases where ingneous aggregates have been used .

Ultrasonic pulse velocity (UPV) measurements may be employed in damage assessment if a sound area o fconcrete is available for comparison and the physical configuration of the member and its reinforcemen tand the surface condition of the concrete are suitable . In complicated cases, laboratory analysis of theresults may be necessary to interpret the readings . Any UPV survey must be preceded by a cover metersurvey to identify the position, orientation and size of steel reinforcement .

Cores may be taken from concrete structures to allow the depth of colour change to be determined or fo rmaterial testing purposes . However, as the damage will be worst at the fire exposed surface, which will b elikely to be in the area of the core nearest the plattens of the testing machine, care must be taken whe ninterpreting the test results . Coring should be carried out in non-critical areas of structures where stresslevels are low .

Thermoluminescence (TL) measurement provides an excellent indication of the thermal exposure thatconcrete has been subjected to and from this the residual compressive strength may be judged . As withUPV testing, however, it must be realised that concrete strengths may occupy a range of values for an ythermal exposure and only an approximate figure may be attributed to the residual strength . TL testing i sparticularly valuable in critical areas of a structure where other methods may not be applied. The TL test i sthe only test that can tell definitely if concrete has been heated significantly and in many cases its mos tuseful application is in indicating areas that have not been subjected to damaging thermal exposure s(especially in cases where a large amount of smoke damage is present) . Post-fire exposure to radiation mayaffect the TL signal .

In prestressed concrete structures, such as containments and pressure vessels, the residual level of prestres swill be of great importance . Installed instrumentation may be used if it has survived the fire and in un-bonded prestressing systems load checks may be carried out and tendons withdrawn for inspection andtesting. In this type of system it is possible to replace and restress tendons that may have been affected b yhigh temperatures in order to restore the level of prestress in the structure. The situation with regard tobonded prestressing systems can be more difficult to assess as withdrawal, replacement and restressingcannot be carried out.

160


Brickwork and Blockwor k

By virtue of their method of manufacture, clay bricks perform well under fire conditions although th emortar used in wall construction and concrete blocks will degrade in a manner similar to that ofunreinforced concrete. Brick and blockwork panels may also be damaged by restrained thermal forces an dby global structural movement of steel or concrete frames and by surface damage due to shock coolingduring fire fighting operations .

Steelwork

In general, a structural member remaining in place, with negligible or minor distortions to the web, flangesor connections should be considered satisfactory for further service . The exception will be for the relativelysmall number of structures in cold-worked or tempered steel where there may be permanent loss o fstrength. The change in strength may be assessed using estimates of the maximum temperatures attained o ron-site tests (such as hardness tests) ; if necessary, the steel should be replaced. Microscopy can be used todetermine changes in microstructure . Since this is a specialised field, the services of a metallurgist areessential [9] . Bolted connections may require detailed consideration if heated above 360°C .

Supplementary Testin g

Supplementary testing such as load testing or detailed materials testing is normally only required in specialcases where the damage assessment is highly critical or where special problems exist . This may includetensile and hardness testing of steel, measurement of the modulus of elasticity of the material, durabilit ytesting or load testing of the entire structure or a portion of it . If a fire has occurred inside a containmentstructure, a structural overpressure test may be required to confirm the structural behaviour andleaktightness of the containment .

Once the residual characteristics of the structure have been determined, a design check based on thematerial properties must be carried out to examine the ability of the structure to continue to fulfil its designrole and to identify areas in need of strengthening .

Further Informatio n

Further detailed information on the effects of fire on structures and structural surveys may be obtained b yconsulting the references given at the end of this document [3-10] .

Table 1 gives details of a typical outline procedure for the post-fire inspection of NPP structures and Tabl e2 gives details of temperature indicators which may be used to assist the investigation .

Recommendation

Although detailed inspection procedures should be written for each structure after a fire event, it i srecommended that an outline higher level generic fire postplan is written which contains a ranking o fnuclear safety related structures (in order of their importance to nuclear safety) and outlining genera lprocedures for post-fire inspection in order to reduce the response time .

161


References :

1.

"The Nuclear Installations Act 1965", HMSO, UK, 1965 (et seq. )

2. McNulty, T, McNair, I, and Bradford, P, "A Regulatory View of Plant Life Management of Civi lStructures in the Nuclear Industry", Institution of Nuclear Engineers, International Conference onNuclear Plant Life Management, 17-19 November 1998, Warrington, U K

3.

Smith, LM, "The Assessment of Fire Damage to Concrete Structures", PhD Thesis, Paisle yCollege of Technology, UK, 1983

4.

Lie, TT, "Fire and Buildings", Applied Science Publishers, London, UK, 197 2

5.

Pchelintsev, VA, (ed), "Fire Resistance of Buildings", Amerind Publishing Co . Pvt Ltd, Delhi,India, 1978

6.

Malhotra, HL, "Design of Fire-resisting Structures", Surrey University Press, UK, 198 2

7.

Schneider, U, "Properties of Materials at High Temperatures - Concrete", RILEM, France, 198 5

8.

Harmathy, TZ, "Fire Safety Design and Concrete", Longman Scientific & Technical, Harlow, UK ,199 3

9.

"Appraisal of Existing Structures (Second Edition)", Institution of Structural Engineers, London ,UK, October 1996

10.

ACI 349 .3R-96, "Evaluation of Existing Nuclear Safety Related Concrete Structures", AmericanConcrete Institute, Detroit, USA, 1996

11.

"Preparation of Fire Hazard Analyses for Nuclear Power Plants", IAEA Safety Reports Series No .8, IAEA, Vienna, Austria, 1998

162


Table 1 Post-Fire Investigation (modified for NPPs) [3]

StageActivity

1 Preparatio n

Obtain (i)

Original design calculations and as-built drawing s(ii)

Plant modification details(iii) Nuclear Safety Cas e(iv) Normal in-service inspection procedures and previous survey reports(v)

Station log(vi) Fire Brigade report s(vii) Eye witness reports(viii) Any available monitoring instrumentation result s

Write

(i)

Specific post-fire inspection procedure(ii)

Specific written system of work/radiological documentation & requirement s2 Visual Examination

(i)

Identify any pre-fire damage (eg settlement )(ii)

Locate seat of fire(iii) Chart progress of fire(iv) Examine debris for maximum temperatur e(v) Locate special damage (eg explosion damage )(vi) Locate areas of above average fire damage(vii) Classify members(viii) Take photograph s

CONCRETE Accuracy q Time q

3 Simple In-situ Testing(i)

Sounding (hammer survey)(ii)

Colour Changes (if applicable) 12-3

AA

4 More Complex In-situ Testin g(i)

UPV Measurement(ii) Cores for colour examination(iii) Deflection measurement

2-42-3

4

B-CB-CB-C

16 3


Table 1 (continued)

5 Laboratory Testin g(i)

Cores for strength(ii) Themoluminescence

(a) For thermal exposure3 C

(b) For residual strength 4 C(iii) Materials testing (other than 5(i)) 2-4 C

(eg Reinforcement & prestressing tendons) 3-4 C-C+

STEELWORK6 In-situ Testin g

(i)

Distortion & Deflectio n(ii) Hardness testing 1-3 A-B(iii) Metallographic investigation 4 A-B

4 C-C+

GENERAL7 Supplementary Testing

(i) In-situ load testing (incl . SOPT & Tendon Load) - C +

(ii) Design check - C +

8 Final Report

Including photographic record, repair recommendations and update to safety case

Post-Fire Investigation Guide (See Table 1 )q Accuracy

Time1 General Guid e2 Limited Accuracy A Relatively quick in-situ examination3 Moderately Accurate B Time consuming in-situ examination4 Accurate C Requires lab analysis and/or time

to recover specimens/analysis

164


Table 2 Post-fire Thermal Indicators [3, 9]

ApproximateTemperature, °C

Indicator & Condition Typical Example Method o fObservation

Class

100+ Paint deteriorates Coatings Visual I, II120 Polystyrene items collapse,

polythene

items

shrivel,PVC degrades

Plastic items, cableinsulation

Visual I

120-140 Polystyrene

softens,polythene softens and melts

Plastic items Visual I

130-200 Polymethyl

methacrylatesoftens

Handles, covers, glazing Visual I

140 Polyurethane foam charre dblack

Thermal insulation Visual I

150 Paint destroyed Coatings Visual I150 PVC fumes Cable insulation Visual I

150-180 Polystyrene melts and flows Plastic items Visual I170 Phenolic resin changes from

yellow to brownWall linings, roof sheets Visual I

180 60Sn-40Pb solder melts Solder joints - electricalequipment

Visual I

200 PVC browns Cable insulation Visual I205 Charring & clay like

appearance of acrylic resinThermal insulation Visual I

250 Charring of wood begins Doors,

floors,

furniturefittings

Visual I

250 Polymethyl methacrylatebubbles

Handles, covers, glazing Visual I

275 Lead base babbitt melts Sliding bearings in pumps& compressors

Visual I

280 Copper instrument tubin gbegins to soften ,recrystallise

Instrument tubing Hardness test II

300-350 Lead, sharp edges rounde dor drops formed

Plumbing, fixtures,shielding

Visual I

390-400 Zinc die casting alloy melts Plumbing fixtures, smallcomponents

Visual I

400-500 PVC chars Cable insulation Visual I420 Zinc coating melts Galvanised steel Visual I

450-870 Austenitic

stainless

stee lsensitised

Piping Metallographic II

480 Asbestos powders/flakes Column packing Visual I540 High temperature scaling

begins on carbon steelsCarbon steel exposed toair

Visual I

595 Bolts tempered to lowerthan normal hardness

ASTM A193 B7 & B1 6bolting

Hardness test II

Indicator Classes - I Unaffected by time (continued)II Events of a more complex nature involving functions of time, temperature & cooling

rate

16 5


Table 2 (continued) Post-fire Thermal Indicators [3, 9 ]

ApproximateTemperature, °C

Indicator & Condition Typical Example Method o fObservation

Class

600-650 Aluminium & alloys melt Small machine parts ,brackets, electricalconduits

Visual I

700-750 Sheet glass softened oradherent

Glazing Visual I

750 Moulded glass rounded Corrugated window glass Visual I760 Gross deformation of low

carbon steelsStructural steel members Visual I

760 Inorganic zinc paintdarkens, spalls off

Structural coatings Visual I

800 Sheet glass rounded Glazing Visual I820 Borosilicate glass softens,

meltsInstrument gauges, sightglasses

Visual I

845 Bolting hardened wellabove normal range

Steelwork connections Hardness test II

850 Sheet glass flowing easily Glazing Visual I900-1000 Leaded red brass/ brass

meltsPlumbing fixtures Visual I

905 Zinc coating boils off Galvanised steel Visual I950 Ni/Au bronze metal melts Thermocouple wave rings Visual I980 Foamglass insulation melts

to a black-grey slagThermal insulation Visual I

1000-1100 Copper melts Wiring Visual I1100-1200 Cast iron melts Castings Visual I

1400 Low carbon steel melts Structural steelwork Visual I

Indicator Classes - I Unaffected by timeII Events of a more complex nature involving functions of time, temperature & cooling

rate

166


Table 3 Classes of Post-fire damage, Characterisation and Description [9 ]

Class Characterisation Description

1 Cosmetic damage, surface Characterised by soot deposits and discolouration. In most cases soot & colourcan be washed off. Uneven distribution of soot deposits may occur . Permanentdiscolouration on high-quality surfaces may lead to rejection . Odour is includedin the class ; it may be difficult to remove but chemicals are available fo relimination.

2 Technical damage,surface

Characterised by damage to surface treatments and coatings . Small extent onlyof concrete spalling or corrosion on uncovered metals. Painted surfaces can berepaired. Plastic coated surfaces need replacement or covering . Minor spallingmay remain or can be re-plastered .

3 Structural damage,surface

Characterised by some concrete cracking and spalling, lightly charred woo dsurfaces, some deformation of metal surfaces or moderate corrosion damage .This type of damage includes Class 2 damage and an be repaired similarly .

4 Structural damage, cross -section (interior)

Characterised by major concrete cracking and spalling, deformed flanges an dwebs of steel beams, partly charred cross sections of timber constructions, an ddegraded plastics . Damage can in many cases be repaired on the existin gstructure. Within the class are also deformations of structures so large that th eloadbearing capacity is reduced, or dimensional alterations prevent proper fittin ginto building. This applies in particular to metal/steel constructions .

5 Structural damage tomembers and components

Characterised by severely damaged structural members and components ,impaired

materials

and

large

deformations .

Concrete

constructions

arecharacterised by extensive spalling, exposed reinforcement and impaire dcompression zone. In steel structures extensive permanent deformations hav earisen due to diminished loadbearing capacity caused by high temperatureconditions. Timber structures may have almost fully charred cross sections .Changes in materials may occur after fire, so they may display unfavourableproperties . Class 5 damage will usually lead to rejection .

167

Unclassified


Organisation de Coopération et de Développement Economique sOrganisation for Economic Co-operation and Development

05-Sep-2002

English - Or. Englis hNUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S

OECD-NEA WORKSHOP ON THE EVALUATION OF DEFECTS ,REPAIR CRITERIA & METHODS OF REPAIR FOR CONCRETESTRUCTURES ON NUCLEAR POWER PLANT S

Hosted by GRS at the DIN Institute in Berlin, German y

10th-11th April, 2002

JT00130932

Document complet disponible sur OLIS dans son format d’origineComplete document available on OLIS in its original format


ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30t hSeptember 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed :

- to achieve the highest sustainable economic growth and employment and a rising standard of living in Membe rcountries, while maintaining financial stability, and thus to contribute to the development of the world economy ;

- to contribute to sound economic expansion in Member as well as non-member countries in the process of economi cdevelopment ; and

- to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance wit hinternational obligations .

The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece ,Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdo mand the United States . The following countries became Members subsequently through accession at the dates indicated hereafter :Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18t hMay 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996), Korea (12thDecember 1996) and the Slovak Republic (14th December 2000) . The Commission of the European Communities takes part in th ework of the OECD (Article 13 of the OECD Convention) .

NUCLEAR ENERGY AGENC Y

The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEECEuropean Nuclear Energy Agency . It received its present designation on 20th April 1972, when Japan became its firstnon-European full Member. NEA membership today consists of 27 OECD Member countries : Australia, Austria, Belgium ,Canada, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg ,Mexico, the Netherlands, Norway, Portugal, Republic of Korea, Spain, Sweden, Switzerland, Turkey, the United Kingdom and th eUnited States . The Commission of the European Communities also takes part in the work of the Agency .

The mission of the NEA is:

- to assist its Member countries in maintaining and further developing, through international co-operation, th escientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclea renergy for peaceful purposes, as well as

- to provide authoritative assessments and to forge common understandings on key issues, as input to governmen tdecisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainabl edevelopment .

Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive wast emanagement, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law andliability, and public information . The NEA Data Bank provides nuclear data and computer program services for participatin gcountries.

In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency i nVienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.

©OECD 200 2Permission to reproduce a portion of this work for non-commercial purposes or classroom use should be obtained through th eCentre français d’exploitation du droit de copie (CCF), 20, rue des Grands-Augustins, 75006 Paris, France, Tel. (33-1) 44 07 4770, Fax (33-1) 46 34 67 19, for every country except the United States . In the United States permission should be obtained throug hthe Copyright Clearance Center, Customer Service, (508)750-8400, 222 Rosewood Drive, Danvers, MA 01923, USA, or CC COnline : http://www.copyright.com/. All other applications for permission to reproduce or translate all or part of this book shouldbe made to OECD Publications, 2, rue André-Pascal, 75775 Paris Cedex 16, France .

2


COMMITTEE ON NUCLEAR REGULATORY ACTIVITIE S

The Committee on Nuclear Regulatory Activities (CNRA) of the OECD Nuclear Energy Agency (NEA) i san international committee made up primarily of senior nuclear regulators . It was set up in 1989 as a forum for theexchange of information and experience among regulatory organisations and for the review of developments whichcould affect regulatory requirements .

The Committee is responsible for the programme of the NEA, concerning the regulation, licensing an dinspection of nuclear installations . The Committee reviews developments which could affect regulatory requirement swith the objective of providing members with an understanding of the motivation for new regulatory requirementsunder consideration and an opportunity to offer suggestions that might improve them or avoid disparities amongMember Countries . In particular, the Committee reviews current practices and operating experience.

The Committee focuses primarily on power reactors and other nuclear installations currently being builtand operated . It also may consider the regulatory implications of new designs of power reactors and other types ofnuclear installations .

In implementing its programme, CNRA establishes co-operative mechanisms with NEA's Committee o nthe Safety of Nuclear Installations (CSNI), responsible for co-ordinating the activities of the Agency concerning th etechnical aspects of design, construction and operation of nuclear installations insofar as they affect the safety of suc hinstallations . It also co-operates with NEA's Committee on Radiation Protection and Public Health (CRPPH) an dNEA's Radioactive Waste Management Committee (RWMC) on matters of common interest .

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION S

The NEA Committee on the Safety of Nuclear Installations (CSNI) is an international committee made upof scientists and engineers . It was set up in 1973 to develop and co-ordinate the activities of the Nuclear EnergyAgency concerning the technical aspects of the design, construction and operation of nuclear installations insofar a sthey affect the safety of such installations . The Committee's purpose is to foster international co-operation in nuclearsafety amongst the OECD Member countries .

CSNI constitutes a forum for the exchange of technical information and for collaboration betwee norganisations which can contribute, from their respective backgrounds in research, development, engineering o rregulation, to these activities and to the definition of its programme of work . It also reviews the state of knowledgeon selected topics of nuclear safety technology and safety assessment, including operating experience . It initiates andconducts programmes identified by these reviews and assessments in order to overcome discrepancies, develo pimprovements and reach international consensus in different projects and International Standard Problems, and assist sin the feedback of the results to participating organisations . Full use is also made of traditional methods of co-operation, such as information exchanges, establishment of working groups and organisation of conferences andspecialist meeting .

The greater part of CSNI's current programme of work is concerned with safety technology of wate rreactors . The principal areas covered are operating experience and the human factor, reactor coolant systembehaviour, various aspects of reactor component integrity, the phenomenology of radioactive releases in reacto raccidents and their confinement, containment performance, risk assessment and severe accidents . The Committeealso studies the safety of the fuel cycle, conducts periodic surveys of reactor safety research programmes and operate san international mechanism for exchanging reports on nuclear power plant incidents .

In implementing its programme, CSNI establishes co-operative mechanisms with NEA's Committee onNuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation ,licensing and inspection of nuclear installations with regard to safety . It also co-operates with NEA's Committee o nRadiation Protection and Public Health and NEA's Radioactive Waste Management Committee on matters ofcommon interest.

3


4


Foreword

The Committee on the Safety of Nuclear Installations (CSNI) of the OECD-NEA co-ordinates the NE Aactivities concerning the technical aspects of design, construction and operation of nuclear installation sinsofar as they affect the safety of such installations . In 1994, the CSNI approved a proposal to set up aTask Group under its Principal Working Group 3 (recently re-named as the Working Group on Integrity o fComponents and Structures (IAGE)) to study the need for a programme of international activities in th earea of concrete structural integrity and ageing and how such a programme could be organised . The taskgroup reviewed national and international activities in the area of ageing of nuclear power plant concretestructures and the relevant activities of other international agencies . A proposal for a CSNI programme ofworkshops was developed to address specific technical issues which were prioritised by OECD-NEA taskgroup into three levels of priority :

First Priority

• Loss of prestressing force in tendons of post-tensioned concrete structure s• In-service inspection techniques for reinforced concrete structures having thick sections and area s

not directly accessible for inspection

Second Priority

• Viability of development of a performance based database• Response of degraded structures (including finite element analysis techniques )

Third Priority

• Instrumentation and monitoring• Repair methods• Criteria for condition assessment

The working group has progressively worked through the priority list developed during the preliminar ystudy carried out by the Task Group . Currently almost all of the three levels of priority are effectivelycomplete, although in doing so the committee has identified other specific items worthy of consideration .By working logically through the list of priorities the committee has maintained a clarity of purpose whic hhas been important in maintaining efficiency and achieving its objectives . The performance of the grouphas been enhanced by the involvement of regulators, operators and technical specialists in both the work o fthe committee and its technical workshops and by liaison and co-operation with complementar ycommittees of other international organisations . The workshop format that has been adopted (based aroun dpresentation of pre-prepared papers or reports followed by open discussion and round-table development o frecommendations) has proved to be an efficient mechanism for the identification of best practice, potentia lshortcomings of current methods and identification of future requirements .

5


6


SUMMARY

OECD-NEA workshop on the evaluation of defects, repair criteria & methods of repair for concret estructures on nuclear power plant s

OECD-NEA IAGE held an international workshop on the evaluation of defects, repair criteria & method sof repair for concrete structures on nuclear power plants in Berlin, Germany on April 10-11, 2002 .Through 2 technical sessions devoted to Operational Experience and State of the Art and Futur eDevelopments, a broad picture of the status was given to a large audience composed by 54 participant sfrom 17 countries and International Organisations . 21 papers have been presented at the Workshop .

The objectives of the workshop were to examine the current practices and the state of the art with regard t othe evaluation of defects, repair criteria and methods of repair for concrete structures on Nuclear Powe rPlants with a view to determining the best practices and identification of shortfalls in the current methods,which are presented in the form of conclusions and recommendations in this report .

This workshop on the evaluation of defects, repair criteria and methods of repair for concrete structures o nNuclear Power Plants is the latest in a series of workshops .

The complete list of CSNI reports, and the text of reports from 1993 on, is available onhttp://www.nea .fr/html/nsd/docs/

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8


Acknowledgement

Gratitude is expressed to GRS, Germany for hosting the Workshop at the DIN Institute in Berlin. Inparticular, special thanks to Mr . Helmut Schulz and Dr Jurgen Sievers, and also Mrs Brunhilde Laue andMrs Schneider for their help .

Thanks are also expressed to chairmen of the sessions and to the Organizing Committee for their effort an dco-operation.

Dr Leslie M Smith BEG(UK) Ltd (UK) ChairmanProf Pierre Labbé IAEA (International )M. Jean-Pierre Touret EdF (France)Herr Rüdiger Danisch Framatome ANP GmbH (Germany)Mr James Costello USNRC (USA)Dr Dan Naus ORNL (USA)M.Eric Mathet OECD-NEA (International)

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10



EVALUATION OF DEFECTS, REPAIR CRITERIA & METHODS OF REPAIR FO RCONCRETE STRUCTURES ON NUCLEAR POWER PLANTS

10th and 11 th April, 2002Berlin, Germany

A. CONTENTSB. CONCLUSIONS AND RECOMMENDATIONSC. PROGRAMMED. PAPER SE. PARTICIPANTS

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12


A. TABLE OF CONTENTS

PAGE

Volume 1

B. CONCLUSIONS AND RECOMMENDATIONSC. PROGRAMMED. PAPER S

Introductory Paper

Inspection, Assessment and Repair of Nuclear Power Plan tConcrete StructuresD. J . Naus, Oak Ridge National Laboratory, U.S.A .H. L. Graves, J.F. Costello, USNRC, U.S.A .

Repair of the Gentilly-1 Concrete Containment Structur eA. Popovic, D . Panesar and M. Elgohary, AECL, Canada

The Repair of Nuclear Power Plant Reinforced Concrete Marine Structures an dInstallation of an Automated Cathodic Protection Syste mL. M. Smith, C . A. Hughes, British Energy Generation UK, Ltd .G. Jones, Sea-Probe, Ltd .

Feasibility Study of IE-SASW Method for the Non-Destructive Evaluation of Containmen tBuilding of Nuclear Power PlantMr. Yong-Pyo Suh, KEPRI, Korea

Field Studies of Effectiveness of Concrete Repair sN.J.R. Baldwin, Mott MacDonald Ltd.,(UK)

Detection and Repair of Defects in the Confinement Structures at Paks NP PMr. Nyaradi Csaba, Paks NPP Ltd

Steam Generator Replacement at Ringhals 3 Containment, Transport Openin gJan Gustavsson, Ringhals Nuclear Power Plant, (Sweden )

In Service Inspection Programme and Long Time Monitoring of Temelin NP PContainment StructuresJan Maly, Jan Stepan, Energoprojekt Prague, Czech Republi c

Repair Criteria and Methods of Repair for Concrete Structures on Nuclear Power Plants

135R. Lasudry, Tractebel Energy Engineering, (Belgium )

SESSION A: OPERATIONAL EXPERIENCEChairman: Mr. Rüdiger Danisch, Framatome-ANP GmbH (Germany)

37

SESSION A: OPERATIONAL EXPERIENCE (Continued )Chairman: Dr James Costello, USNRC (USA)

115

171923

25

39

5 1

59

73

83

117

125

13



157L.M. Smith, British Energy Generation UK, Ltd ., (UK)

Volume 2

SESSION B: STATE OF THE ART & FUTURE DEVELOPMENTS

17Chairman: Dr Naus, ORNL (US)

Various stages to Address Concrete Cracking on NPP sC. Seni, Mattec Engineering Ltd ., (Canada)

Investigation of the Leakage Behaviour of Reinforced Concrete Wall sNico Herrmann, Christoph Niklasch, Michael Stegemann, Lothar Stempniewski ,University of Karlsruhe, (Germany )

The Development of a State-of-the-Art Structural Monitoring Instrumentation Syste mfor Nuclear Power Plant Concrete StructuresL. M. Smith, B . Stafford, M.W. Roberts, British Energy Generation UK ,Ltd.A. McGown, University of Strathclyde (UK)

Ageing and Static Reliability of Concrete Structures under Temperatur eand Force LoadingPetr Stepanek,, Stanislav Stastnik Vlastislav Salajka, Technical University of Brno ,Jaroslav Skolai, Jiri Stastny, Dukovany Power Plant, (Czech Republic )

Efficient Management of Inspection and Monitoring Data for a Bette rMaintenance of InfrastructureMarcel de Wit, Gilles Hovhanessian, Advitam

Aging Process Of A Good Concrete During Forty Year sDr. Peter Lenkei, University, College of Engineering (Hungary)


8 1Chairman: Mr. Jean-Pierre TOURET, EdF, (France)

Acoustic Monitorin gMarcel de Wit, Gilles Hovhanessian, Advitam

Concrete Properties Influenced by Radiation Dose During Reactor Operatio nTakaaki Konno, Secretariat of Nuclear Safety Commission, (Japan )

The Use of Composite Materials in the Prevention and Strengthening of Nuclea rConcrete StructuresD. Chauvel, P .A. Naze, J-P . Touret EdF, Villeurbanne, (France)

19

3 1

4 1

55

67

77

83

97

105

14


Detection of Reinforcement Corrosion and its Use for Service Life Assessment o fConcrete StructuresC. Andrade, I . Martinez, J . Munoz, CSIC (SP) Rodr iguez, M. Ramirez, GEOCISA (Spain)

Improved Detection of Tendon Ducts and Defects in Concrete Structures Usin gUltrasonic Imagin gW. Müller, V. Schmitz, FIZP, (GE) M. Krause, M . Wiggenhauser, Bundesanstalt fürMaterialforschung und -Prüfung, (Germany )

Structural Integrity Evaluation of a Steel Containment for the Replacement o fSteam GeneratorMr. Yong-Pyo Suh, KEPRI (Korea)

New Methods on Reconstruction of Safety Compartments of Nuclear Power Plants

141Z. Kdpper, Kdpper und Partner, Bochum, (Germany)D. Busch, RWE Solutions AG, Essen, (Germany)

E.

PARTICIPANTS

151

115

125

133

15


16


SESSION B: STATE OF THE ART & FUTURE DEVELOPMENT SChairman: Dr. Naus, ORNL (US )

17


18


Addressing concrete cracking in NPPs.

By C. Seni ,Mattec Engineering Ltd. Canada

Abstract

The phenomenon of concrete cracking is one of the most frequently encountered deterioration at NPPs as i thas been shown by a wide Survey of NPPs performed by IAEA in 1994-9 5

It can be due to a multitude of causes such as the normal ageing process (shrinkage, creep, prestressin gforce loss) as well as exposure to the environment (temperature variation, moisture, freeze/thaw, etc )

The above mentioned Survey has also shown that in 64% of cases, no action was taken or required . Itbecame also obvious that there is a lack of guidance as when remedial actions are needed .

The paper describes, with the help of a Flow Chart, the various stages to be considered, from the first stepof identification of cracks, to the definition of causes, evaluation of extent of damage, evaluation o feffect/implications (safety, reliability) , to the final step of deciding if repair action is required .

Finally, based upon a wide literature survey the paper proposes in a Chart format, Criteria for addressingconcrete cracks in NPPs ., when taking in considerations all these factors .

General

Reinforced concrete structures deteriorate in various ways due to the normal ageing process ( shrinkage ,creep, prestressing cables relaxation) and/or impact from aggressive environment (temperature, moisture ,cyclic loading )

The rate of deterioration will depend on the component’s design, material selection, quality of construction ,curing, and aggressiveness of the environment .

The experience gained from an international survey on ageing of Nuclear Power Plants initiated by IAE Ain 1994-95 [2] was that concrete cracking was the most frequent form of degradation . At the same time theSurvey has shown that in 64% of cases no action was required .

Various reasons could explain this lack of action which could have unpleasant implications .

In-service inspection techniques are available that can indicate the occurrence and extent of such an agein gor environment-stressor related deterioration. Periodic application of these techniques as part of acondition assessment program can indicate the progress of deterioration . The results obtained from thes eprograms can be used to develop and implement remedial actions before the structure attains anunacceptable level of performance . Depending on the degree of deterioration and the residual strength o fthe concrete component, the remedial measures may be structural, protective, cosmetic, or any combinatio nof these .

19


A plant service life management program would normally include measures for detection and monitorin g(inspection, instrumentation, reporting, mapping, etc,), assessment of extent of damage (measurement) andtype of damage (impacting safety, reliability), technicalities (possibility to repair, cost, duration) ,scheduling (prioritization), and selection of a repair method .

However the world literature as well as codes and standards ,if not confusing, at least are lacking clea rindications when to proceed with repair, taking in consideration all aspects involved .

This paper discusses the process which should lead to the selection of an effective repair method andproposes, based upon worldwide standards and literature, criteria which should lead to the decisionwhether to repair or not concrete cracks, after the cracks have been identified and evaluated, addressing th eentire range of aspects involved .

Lead way to the selection of a repair metho d

Definition of Cause

Once cracks have been identified through periodic inspection or instrumentation, their location anddimension recorded and evolution monitored, the next step is the definition of causes, since no repai rshould be undertaken before the cause of cracks has been identified .

Concrete ageing, external factors or simply the aspect of the cracks could provide information abou tcauses. In each of these three categories there are a number of specific indicators (Chart#1) .

(i) Cracks due to concrete ageing .

One category of cracks is the result of the ageing phenomenon of the concrete per se, and with time crack swill multiply or increase in width, depth or length, thus giving an indication of the status of ageing of tha tconcrete .

The reasons for cracking which are time dependent could be :

- shrinkage/creep- prestressing loss of the cables- reinforcing bars corrosion .

The creep of the concrete will be present in particular in prestressed concrete structures and the associate dloss of prestressing force viz cable relaxation should always be considered as a potential cause of cracking,while this cause will not be present in conventionally reinforced concrete .

(ii) Cracks due to external factors .

Another category of cracks is the result of the various external factors independent of the concrete agein gbut associated with ageing since they have an acceleration effect upon it .In this category fall cracks resulting from the effect of exposure to the environment :

-freeze/thaw cycle,-chloride penetration,-carbonation,-aggressive environment (e .g .atmospheric pollutants, acid rain, fog, sea/ground water exposure. )

20


-cycling loading ( e .g . mechanical/thermal),-construction defects (e .g.low concrete quality, excessive permeability, alkali reactive aggregates-AAR, early formwork removal ) ,

-design criteria (allowing for concrete in plastic/non-linear state, thus with acceptable crackformation in the tension zone, insufficient knowledge of some long term acting desig nparameters) ,

-thermal gradient-accidents (e .g . loss of coolant-LOCA, fire, earthquake)-excessive testing, (e .g . containment leak rate test )

(iii) Cracks aspectThe location or pattern (direction) off cracks could also provide indication about the reason of thei rformation.

To recognize, viz differentiate between all these causes and heir related effect, the world literature and i nparticular the extensive work and publications produced by RILEM and IAEA can provide ampl einformation . [1], [3], [4], [5], [6], [7], [8], [13] .

Extent of Damage.

Two aspects need to be considered in the process of identification of the extent of damage i .e . the cracksdimension and their evolution .(Chart #1) .

(i)Cracks dimension .

There are three parameters which need consideration, i .e . the width, length and depth of the cracks . Theseparameters are usually monitored and IAEA Survey [2] has shown that, world wide ,- the crack width was recorded by 73 .3% of the stations,- the length was recorded by 74% of the stations ,- the depth was recorded by 21 .4% of the stations .

Recording crack width and/or depth is important since the crack size can affect the corrosion of thereinforcing, which in turn will amplify the effect of the freeze/thaw cycles and the ingress of otheraggressive agents (e .g. chlorides) to the proximity of the reinforcing bars . Various authors have rated thesignificance of the crack width based upon these aspects as shown in Table1, [3], [7] . From Table1 itappears that crack widths between 0 .1mm and 0.4 mm are considered acceptable, depending on th eenvironmental conditions .However it does not address the complexity of the cracking phenomenon with all aspects involved .

(ii)Cracks activity

Recording crack length is important in order to determine their evolution, since this will indicate if cracksare active, i .e . the process accountable for the cracks is still in effect and will have to be identified an daddressed, while no change in the crack length ( dormant cracks) could indicate that the process ha sstopped and may not be age related.

There are one time cracks like those due to an accidental loading or construction defects and which will no tfurther develop, and cracks originated by time dependent factors like concrete “ageing per se” (e .g.carbonation) or environmental factors (e .g . chlorides, freeze/thaw) .

21


To the first category belong the Dormant cracks while to the second category belong the Active crack s(Table2), [3], [8] .

This classification is important when assessing the urgency for repair or the repair method . Also in the cas eof Active cracks the cause should be first eliminated, before a repair is undertaken .

Table 2 however is limited to a few cases and can not provide sufficient guidance .

Impact of Damage.

The next step of the process consists of establishing the impact the cracks can have upon the concret emember in particular and upon the structure in general, since this will affect the decision whether t oproceed or not with repairs . This step involves consideration of the following ;

Structural member ranking.The effect of cracks will depend upon the structural member ranking which should take into account th eimportance and function of the member based upon its structural and functional role which will indicate t owhat extent the safety or functional performance are affected . Finally this will lead to the necessity andurgency to proceed with repair s

A few methodologies were suggested for ranking nuclear plant components, i .e . in the US according to aprogram elaborated at the Oak Ridge National Laborator y

[7],[9] ,or in the UK [10], or in Canada [11], and by RILEM [8] .These should provide the basis for each NPP to develop their own methodology .

Assessment .

Before selecting a type of repair, the last step in the process is the assessment of the necessity to proceedwith repairs or not . This should be based upon a Cracks Acceptance Criteria

The best attempt so far was made by C .J . Hookham in 1995 taking in consideration the crack dimensionand environmental factors (chloride penetration and depth of carbonation), as shown in Chart [6] .

Going with the complexity of the crack phenomenon one step farther is the Cracks Acceptance Criteri apresented in this paper and as shown in a chart form in Figure 1 and 2 .The two charts include the following parameters for consideration :

-crack width (range 0 .2 mm to 1 .0 mm) ,-type of environment (mild or aggressive) ,-cracks activity (active or dormant), and-depth of chloride penetration or carbonation (low or high) .

The values set are based upon a review of the world literature and codes and standards referenced in thi spaper.

The crack width is closely related to the urgency of repair since this represents an open path for th eaggressive agents to reach the reinforcement .

22


The upper limit is 0 .2mm beyond which repairs are not required under any circumstances while beyond th elower limit of 1 .0 mm, repairs are required in all cases . For in-between values indication are given in thetwo charts .

The aggressiveness of the environment will also affect the decision . Thus an aggressiveenvironment ( e .g . sea shore location with high chloride content, proximity to air pollutant industries, hig hfreeze/thaw cycle, etc ) will increase the limit and associated repair urgency factor .

Regarding the cracks activity, for the dormant ones the necessity of repair will be less stringent a sreflected in the two charts .

The depth of chloride penetration and/or carbonation are indicators of how close the initiation o freinforcement corrosion is . Maximum permissible chloride contents, as well as minimum recommendedreinforcement cover requirements have been provided in codes and guides . The threshold of acid-solubl echloride contents reported by various investigators which could initiate steel corrosion ranges from 0 .15 to1 .0% by weight of cementitious materials, whereas code limits range from 0 .2 to 0 .4%. [3],[7],[13] . In thetwo charts (Fig.1 and 2) the limit selected is 0 .4% .

As earlier indicated the cracks have also to be considered in connection with the importance of th estructural member affected which will come from the Plant Components Ranking . For a high rankingcomponent (Figure 1) the urgency of repairs will be grater than for a low ranking one (Figure 2) .

Recommendations.

Each NPP should develop a concrete component ranking as well as a database (history) of crack repair sperformed and their effectiveness .

An experienced civil engineer familiar with concrete ageing, should be associated with the entir eassessment process described.

Repair should proceed only after the cause has been determined and the repair should include th eelimination of cause .

Concluding remarks

The paper has detailed the various steps to follow when dealing with concrete cracks, from the time of thei ridentification to the decision making of when and how to proceed with their repair, with Chart#1 as a guid ethrough the various steps involved .

Figure 1 and 2 are the Acceptance Criteria to follow at the end of the assessment process when all requiredcrack parameters are known, to provide the answer if and when remedial action is required .

The selection of repair materials and method were not within the scope of this paper .

However such information, when required, can be found in the world literature or from the experience o fother similar NPPs using inter communication grids like WANO or COG since most Utilities will have adatabase of repairs performed and their effectiveness[11] .

23


REFERENCES.

1 . A.M.Neville, Properties of concrete, Pitman Publishing .

2. Summary Results of the Survey on Concrete Containment Ageing, IAEA/NENS Working Materia l,

Vienna, 199 5

3. Assessment and management of ageing of major NPP components important to safety : Concretecontainment buildings, IAEA-TECDOC-1025, June 199 8

4. L.Granger, Assessment of creep methodologies for predicting prestressing forces losses in nuclea rpower plant containment, -RILEM Report 19/1999 , Considerations for use in managing the agein gof NPP concrete structures .

5. Assessment and management of ageing of major NPP components important to safety : Metalcomponents of BWR containment systems, IAEA-TECDOC-1181, October 200 0

6. C.J . Hookham, In-service inspection guidelines for concrete structures in NPPs -ORNL/NRC/LTR -90 Lockneed Martin Energy Systems Inc .,ORNL,Oak Ridge,1995

7. D.J . Naus/C.J .Hookham, Condition assessment of concretestructures in NPPs, RILEM Repor t19/1999 , Considerations for use in managing the ageing of NPP concrete structure s

8. D.J .Naus/ P .D. Krauss/C.Seni, Repair techniques and materials for degraded NPP concret estructures, -RILEM Report 19/1999 , Considerations for use in managing the ageing of NP Pconcrete structures .

9. C.J . Hookham, Structural ageing assessment methodology for concrete structures in NPPs ,ORNL/NRC/LTR-90/17 Martin Marietta Energy Systems Inc .,ORNL,Oak Ridge,Tenn.March 199 1

10. R.Judge, Classification of structural components and degradation mechanisms for containmen tsystems, -Proceedings of the Third International Conference on Containment Design and Operation ,Toronto, Oct .1994

11. K.E. Philipose, The structural aging assessment program-Ranking methodology application forCandu NPP concrete components, COG-97-269, May 199 7

12. J.A. Sato et . al , Effective repairs in concrete containment -Materials and guidelines for using join tsealant, patching concrete and repairing cracks in CANDU containmen t

structures, COG-96-50 1

13. C.Seni, Methodology for Assessment of Repair of Concrete Cracks and Acceptance Criteria, COG -97-202, Decmber 1997

24


Table 1: Permissible crack widths to prevent corrosion of steel reinforcement . [3],[7 ]

Author Environment factors Permissible width, m m

RengersDangerous crack width 1 .0 to 2 . 0

Crack width alowing corrosion within 1/2 yea rsaline environment

0 . 3

Abeled Structures not exposed to chemical influences 0 .3 to 0 . 4

Tremper Found no direct relation between crack width and corrosio n

Boscard Structures exposed to a marine environment 0. 4

de Bruyn Found no direct relation between crack width and corrosio n

Engel and Leeuwen Unprotected structures (external) 0. 2

Protected structures (internal) 0 . 3

Voelmy Safe crack width up to 0 . 2

Crack allowing slight corrosion 0.2 to 0 . 5

Dangerous crack width over 0 . 5

BerteroIndoor structures 0 .25 to 0 .3 5

Normal outdoor exposure 0 .15 to 0 .2 5

Exposure to sea water 0 .025 to 0 .1 5

HaasProtected structures (interior) 0 . 3

Exposed structures (exterior) 0 . 2

BriceFairly harmless crack width 0 . 1

Harmful crack width 0. 2

Very harmful crack width 0. 3

SalingerFor all structures under normal conditions 0. 2

Structures exposed to humidity or to harmful chemical influences 0. 1

WastlundStructures subjected to dead load plus half the live load for which the yare designed

0 . 4

Structures subject to deal load only 0 . 3

EfsenExterior (outdoor) structures exposed to attack by sea water and fumes 0.05 to 0 .2 5

Exterior (outdoor) structures under normal conditions 0 .15 to 0 .2 5

Interior (indoor) structures 0 .25 to 0 .3 5

RuschOrdinary structures 0 . 3

Structures subjected to the action of fumes and sea environment 0.2

25


Table 2 : General guide to repair options for concrete cracking . [3],[8 ]

Description Repair OptionsPerceived

durability Commentaryrating (1-5*)

Dormant pattern Judicious neglect 4 Only for fine cracksor fine cracking Autogenous healing 3 Only on new concrete

Penetrating sealers 2 Use penetrating sealer for H20,

Coatings 3Cl resistanceUse coating for abrasion and

HMWM or epoxy treatment 2chemical resistanceTopical application, bonds

Overlay or membrane 2cracksFor severely cracked areas

Dormant Epoxy injection 1 Needs experienced applicatorisolated

large Rout and seal 3 Requires maintenanc ecracking Flexible sealing 4 Requires maintenanc e

Drilling and plugging 3Grout injection or dry 4packing 5Stitching 4Additional reinforcing 3Strengthening

Active cracks Penetrating sealer 3 Cracks less than 0 .5 mmFlexible sealing 3 Requires maintenanc eRoute and seal 3 Use for wide cracksInstall expansion joint 2 ExpensiveDrilling and plugging 4 May cause new cracksStitching 4 May cause new cracksAdditional reinforcing 3 May cause new cracks

Seepage Eliminate moisture source 2 Usually not possibleChemical grouting 1 Several applications may be

Coatings 4necessaryMay have continued seepag e

Hydraulic Cement dry 4 May have continued seepag epackaging

* Scale from 1 to 5, with 1 being most durable .

26


No visible degradation I

Concrete cracking with or

I Concrete cracking,or cracking

w/o staining

staining, and spalling0 < w < 0.4 mm

0 .4 < w < 1 .0 mm

w > 1.0 mm

0 .4 < Cl < 1or

Carbonation DepthApproaching Stee l

FIG. 7.3 . Damage state chart relating environmental exposure, crack width ,and necessity for additional evaluation or repair [Ref. 7 .6] .

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29


30


31


32


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34


35

36


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40


The Development of a State-of-the-Art Structural Monitoring Instrumentation System for Nuclea rPower Plant Concrete Structure s

LM Smith British Energy Generation(UK)Lt dB Stafford British Energy Generation(UK)Lt dMW Roberts British Energy Generation Ltd

A McGown University of Strathclyde

Abstract

This paper describes the development of a state-of-the-art diverse monitoring system for application toexisting concrete structures on Nuclear Power Plants . The system has the capability to monitor surfacebehaviour or, in cases where surface effects are not the most critical, sub-surface behaviour can b emonitored with minor modification to the installation arrangement .

The system uses instrument clusters with both fibre optic and AC-LVDT transducers designed to monito rsmall structural displacements accurately . By simultaneously measuring the structural response usin gtransducers of different types, type-based errors may be eliminated and the system reliability enhanced .

In order to determine the type of instrumentation to be used, a research project was undertaken to evaluat ethe performance of available equipment and the practicality of its installation. This paper describes thework carried out under the research project and the development of the system to the practical installationstage .

Introduction

Historically, emphasis has been placed on the use of instrumentation to validate design and analysi sassumptions and for initial structural integrity testing and it has been normal for the first structure of a newdesign or series to be extensively instrumented with subsequent structures receiving less attention.Although instrumentation has been used for long term monitoring this has often been as a result of thecontinued operation of systems installed for commissioning and structural integrity tests, which have thenbeen adopted for long term monitoring purposes . It is now generally accepted that the installation o fstructural monitoring systems at the time of construction will provide useful information for the lifetim emanagement of nuclear power plant structures and the detection of ageing effects .

The OECD-NEA workshop on the instrumentation and monitoring of concrete structures [1] consideredthis in some detail . There is now a perceived need to address the ageing of concrete structures . To this end,the use and acceptance of instrumentation techniques has increased with time and more reliance is bein gplaced on such techniques .

The workshop recommended that, while techniques are available to monitor the ageing and performance o fstructures and new systems are available which may be retrofitted, improvements be made in retrofittinginstrumentation and in relating it to existing instrumentation . It is extremely difficult to replaceinstrumentation that was installed at the time of construction . Additionally, the usefulness of installedinstrumentation is dependent on the accuracy and reliability of the sensors used . The usefulness ofinstrumentation systems has also been improved by developments with regard to computer managemen tsystems and databases .

41


The Current Project

British Energy identified a potential need to be able to measure small changes in structural displacement inreinforced and, particularly, prestressed concrete structures using retro-fitted instrumentation . This paperdescribes the development of a state-of-the-art diverse monitoring system for application to existingconcrete structures on Nuclear Power Plants . The system has the capability to monitor surface behaviou ror, in cases where surface effects are undesirable, sub-surface behaviour with minor modification to th einstallation arrangement. The specification required the investigation of an instrumentation package chose nto include established and innovative technologies, and instrumentation diversity and redundancy .

This paper describes the development of a state-of-the-art diverse monitoring system for application toexisting concrete structures on Nuclear Power Plants . The system has the capability to monitor surfacebehaviour or, in cases where surface effects are not the most critical, sub-surface behaviour can b emonitoredwith minor modification to the installation arrangement .

Programme of Investigation s

The project was initiated in April 2000 and was undertaken at the University of Strathclyde in three stage swith project milestones as follows :

Stage 1

Investigation and definition of the monitoring requirements . Preparation of proposals for theinstrumentation required. This work was completed by mid - May 2000 .

Stage 2

Laboratory investigations at the University of Strathclyde and preparation of an Interim Report o nthe outcome of the investigations . This work was completed in late-December 2000 and an Interi mReport was issued in January 2001 . A recommendation of the Stage 2 work was that an extension toStage 2 be granted to investigate the installation problems associated with the use of Automati cCrackmeters penetrating down below the surface of the concrete some 50 to 100 mm .

Stage 2 (extension )

The Stage 2 (extension) was undertaken during April and May 2001 . This work involved developingtwo types of Automatic Crackmeters, installing them in specially constructed reinforced concret ebeams and testing their measurement efficiency .

Stage 3

The Final Report set out the performance and accuracy of the instrumentation and includedrecommendations for the instrumentation system to be used on NPPs .

42


Results

In Stage 1 of this project the normal operational conditions on Nuclear Power Plants were assessed and th emost likely types of instrumentation to effectively undertake the monitoring were identified . In Stages 2and 2(extension), evaluation of these instrument types and the methods of installing these wer einvestigated (Plate 1) . Stage 3 involved the preparation and submission of the Final Report .

The Stage 2 investigations were designed to measure vertical distortion, in-plane deformations / strains an dtemperatures on a centrally loaded, simply supported 3m x 1m x 100mm lightly reinforced concrete beam .The instrumentation package was chosen to include established and innovative technologies ,instrumentation diversity and redundancy . The range of instrumentation employed is shown in Table 1 .The data on the claimed resolution, frequency of measurement, mounting and development stage of eachinstrument are shown in Table 2 . The details of the installation methodology, tuning/ commissioning an ddata collection technique associated with these instruments are given in Table 3 . The layout of theinstrumentation employed on the beam is given in Fig .1 and illustrated in Plate 1 . In addition to thes einstruments, ambient air temperature was measured using PRT’s above, below and at the side of the beam .The objectives and details of the testing regime are shown in Table 4 . Incremental vertical distortion tes tdata was obtained from two separate tests, involving four stages of short-term loading with a single stageof unloading. In addition, tests were conducted to compare the analogue and digital outputs from the fibreoptic (Fabry-Pérot interferometer) instruments . The reason for undertaking these tests was to prove thatanalogue signals could be obtained from the fibre optic instruments, as this had not been establishedpreviously. The main findings of the testing undertaken under Stage 2 were :

Vertical and in-plane deformations could be measured by the instrumentation employed to anaccuracy of +/- 0 .01mm, (+/- 10 microns) .

In-plane strains could be measured to an accuracy of +/- 2 microstrain .

Temperature could be measured to an accuracy of +/- 0 .1 o C .

AC- LVDT’s and fibre optic deformation gauges were the preferred instruments to measur edeformations .

Whilst highly accurate and possibly extremely useful in the future, some further development wor kwas required on the TMS laser system before it could be employed for structural monitoring .

VW strain gauges and fibre optic strain gauges were the preferred instruments for directlymeasuring strain .

Digital outputs from the instrumentation signal conditioners were preferred as they allowed muc hgreater flexibility in data collection and networking . However, it was shown that analogue output smay be obtained with similar accuracies to digital outputs .

Instrumentation should not be surface mounted to avoid superficial cracking and de-bondingproblems .

The use of Automatic Crackmeters should be considered to measure in-plane deformations /strains . The pillars and instrument fixings should be manufactured from a temperaturecompensated alloy such as Invar (Fe/Ni alloy) to reduce potential temperature effects .

43


The temperature variations at the location of each instrument should be measured and temperatur ecompensations / corrections applied .

It was recognised that potential installation problems have a major influence on the choice ofinstrumentation and consequently on the measurements to be made . Indeed, it was realized that the need tocreate a horizontal reference plane against which out of plane distortion could be measured, was notpractical at many NPP locations . Thus measurement of out of plane distortion was not recommended an dthe choice of instrumentation was limited to those instruments that could measure in-plane deformation sand strains .

It was identified that superficial cracking of the concrete, free surface effects and de-bonding o finstruments on the surface of the concrete could result in mis-leading data being collected . Thus it wa sagreed that the instrumentation packages to be used would require to be capable of measurement at a sub-surface level . It was recommended that sub-surface deformations and strains should be measured usingboth conventional electrical and optical fibre instruments . It was decided that measurements should bemade at a depth of 50 to 100mm below the top surface of the concrete to avoid surface effects influencingthe data collected .

The study recommended that sub-surface deformations should be measured using Automatic Crackmeter sconsisting of pillars drilled and fixed into the concrete with the change in distance between pairs of pillar smeasured using external AC electrical LVDT’s and fibre optic deformation gauges . The types ofinstruments should be similar to those used in the Stage 2 investigations, however, they should b eindividually temperature monitored to check for local temperature variations on the top surface of th econcrete .

Sub-surface strains should be measured using temperature compensated VW strain gauges and optical fibr estrain gauges fixed within Automatic Crackmeters .

Ambient air temperatures immediately above the concrete surface and concrete temperatures at 25 to50mm below the surface should be measured.

Although the correlation between analogue and digital outputs from the signal conditioners was good, itwas recommended that digital outputs should be the preferred form as :

Links between instruments and signal conditioners were simpler .Site specific problems of mutual interference and external interference were avoided .Networking of instrumentation using RS422 / RS485 links was possible .

The Stage 2 (extension) investigations were designed to measure in-plane deformations / strains an dtemperatures on a centrally loaded, simply supported 3m x 1m x 150mm reinforced concrete beam. Theinstrumentation packages used were designed to provide examples of possible combinations of establishe dand innovative technologies with instrumentation diversity and redundancy .

The two instrumentation packages used were :

A four instrument package, consisting of a VW strain gauge (subsurface), a fibre optic strain gaug e(subsurface), an AC LVDT (external) and a fibre optic deformation gauge (external), and ,

A two instrument package, consisting of an AC LVDT (external) and a fibre optic deformationgauge (external) .

44


The arrangements of the instrumentation packages are shown in Plates 2 and 3 and the layout of th einstrumentation was as shown in Figure 2 .

Test results were obtained from a nine stage loading and single stage unloading test undertaken over seve ndays . These indicated that for lightly loaded and unloaded conditions there were identifiable variations o fthe outputs, (“instrumentation noise”), of +/- 0.002 mm (+/- 2 microns) and +/- 1 microstrain . In addition,creep of the concrete could be identified over time . Thus the main findings of the testing undertaken were :

The two types of instrumentation packages could be installed on the beam efficiently .

Owing to the use of subsurface instruments, the installation of the four-instrument packageinvolved a great deal more drilling than the two-instrument package where the instruments ar eexternal to the concrete surface and drilling is only required for the fixing posts . .

The detailed design of the instrument packages and the adjustment /calibration of the instrument srequired great care to ensure that the instruments were not damaged during installation an doperated efficiently in place .

To calculate the strains in the concrete from the LVDTs and fibre optic deformation gauge outputdata, the gauge length was taken as the distance centre to centre between the vertical pillars .

To calculate the strains in the concrete from the VW and fibre optic strain gauge output data, th egauge length was taken as the distance centre to centre between the vertical pillars and the strain srecalculated on this basis. All data were efficiently recorded using digital outputs, ( i .e . using therecommended / preferred method of Stage 2) .

In plane deformations could be measured by the LVDTs and fibre optic deformation gauges to a naccuracy of +/- 0 .01 mm, (+/- 10 microns) .

Background noise in the LVDTs and fibre optic deformation gauges was at a level of +/- 0 .002mm, (+/- 2 microns) .

In plane strains could be measured by the VW and fibre optic strain gauges to an accuracy of +/- 2microstrain .

Background noise in the VW and fibre optic strain gauges was at a level of +/- 1 microstrain .

Creep occurring in the concrete over time was reflected in a gradual change in the outputs from th einstrumentation .

Final Recommendations

Instrument packages consisting of a VW strain gauge, a fibre optic strain gauge, an AC-LVDT and a fibreoptic deformation gauge should be used . The strain gauges should be temperature compensated and beembedded into the concrete to a depth of 50 to 100mm using a shrinkage compensated cementitious groutand an epoxy “jacket” to isolate the instruments from local effects . The AC-LVDT and fibre opticdeformation gauge should be mounted above the surface, PRTs should be embedded at 25 and 50 mm t omeasure concrete temperatures . PRTs should be mounted next to the bodies of the AC-LVDTs and fibr eoptic deformation gauges to measure local temperatures and so allow temperature corrections to be made .

45


It should be noted that the instrumentation package gauge length must be precisely determined to allow th eeffective gauge length of the strain gauges to be calculated and so their strain outputs corrected . Theinstrumentation package gauge length is used to calculate strains from the LVDT and deformation gaugeoutputs .

An alternative reduced package may be used where space is restricted. This should consist of an ACelectrical LVDT and a fibre optic deformation gauge placed one above the other between two vertica lpillars . The instruments should be placed just above the top of the concrete surface . The vertical pillar sshould be embedded into the concrete to a depth of 50 to 100mm . PRTs should be embedded at 25 and 5 0mm to measure in concrete temperatures and a PRT should be mounted next to the bodies of the AC-LVDT and fibre optic deformation gauge to measure local temperatures and so allow temperatur ecorrections to be made. It should be noted that the instrumentation package gauge length must be preciselydetermined to allow calculation of the strains from the AC-LVDT and deformation gauge outputs .

Specified types of VW and fibre optic strain gauges, AC-LVDTs, fibre optic deformation gauges and PRT swhich have been proven to be accurate and reliable for the purpose of monitoring deformations and strain sat the very low levels required should be employed in the instrumentation packages .

The installation method should be very carefully specified to provide holes in the concrete into which th einstrumentation packages will fit. Thus close tolerances should be specified on hole sizes, spacing an dorientation .

The instrumentation packages should be designed and constructed to close tolerances to ensure :

The instruments are not damaged during fabrication .The instruments can be calibrated before installation and, at least for the AC-LVDTs an ddeformation gauges, after installation .The instruments can be easily set to any specified point in their range prior to installation .

The grouting of the instrumentation packages into the concrete should be very carefully specified an dcontrolled . (The size of the holes in relation to the size of the vertical posts must take account of th egrouting process) .

The instrument package and associated cabling should be very carefully protected from damage afterinstallation .

Calibration of the instrumentation should be very precisely undertaken at a range of movement and rang eof temperatures appropriate to the operational conditions expected in service .

The output from the instrument signal conditioners should be digital . This provides much greater flexibilit yin data collection, allows networking and avoids site-specific interference problems . The use of theanalogue signals is possible if other factors preclude the use of digital signals . The signal conditioners to b eused should be those specifically designed by the manufacturers of the instrumentation for the variou sinstruments . They should be connected to the instruments with the appropriate cable type and length . Thecabling should have the minimum possible number of joints/connections .

The conditioned digital signals should be collected and processed in a PC using specialist data collectio nand processing software with the capacity to collect, process and store data at an appropriate rate for aspecified period . The PC thus requires to possess sufficient processing and storage capacity.

46


“Instrumentation noise” and evidence of creep in the concrete over time, are likely to be a feature of theoutputs from all the instruments . Thus it will be necessary to monitor structures in order to identify thei r“normal engineering behaviour” .

In-service Performanc e

A number of the two-instrument external instrument packages have been installed on actual NPP structure sand these are giving very good results with levels of noise and accuracy comparable with the laboratorytests .

References

1 . OECD Nuclear Energy Agency, NEA/CSNI/R(2000)15 “Proceedings of the Workshop onInstrumentation & Monitoring of Concrete Structures 22-23 March 2000, Brussels, Belgium .”

47


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54


Ageing and Static Reliability of Concrete Structures under Temperature and Mechanical Loadin g

Petr Stépanek a, Stanislav 9stni d , Vlastislav Salajka, b , Petr Hradil b, Jaroslav Sol , c, Jifi9fastnÿ c

a Department of Concrete and Masonry Structures, Technical University of Brno, Ùdolni 53, 60200 Brno ,Czech Republic, e-mail : stepanek .p@fce .vutbr.czb Department of Building Mechanics, Technical University of Brno, Ùdolni 53, 60200 Brno, Czec hRepublic, e-mail : salajka [email protected] Power Plant Dukovany, 675 50 Dukovany, Czech Republi cd Department of Technology and Building Materials, Technical University of Brno, Ùdolni 53, 6020 0Brno, Czech Republic, e-mail : [email protected] .cz

Abstract

The contribution presents some aspects of the static reliability of concrete structures under temperatur eeffects and under mechanical loading. The mathematical model of a load-bearing concrete structure wa sperformed using the FEM method . The temperature field and static stress that generated states of stres swere taken into account. A brief description of some aspects of evaluation of the reliability within th eprimary circuit concrete structures is stated . The knowledge of actual physical and mechanicalcharacteristics and chemical composition of concrete were necessary for obtaining correct results ofnumerical analysis .

Studied problems were divided into a number of fields and worked out in details :

1. Verification of contemporary physical and mechanical characteristics of concrete (input parameters o fthe FEM models) .

2. Checking of the concrete microstructure and verification of the grade and kind of possibl emicrostructure changes .

3. Experimental verification of the boundary conditions from point of view of the temperature field andthe radiation stress .

4. Setting up a mathematical model of the structure for an examination of the interaction of temperatur eand static stresses (finite element method, software ANSYS) in two alternatives :

a) Macro-model representing the essential part of concrete structures in the proximity of the reactor ,b) Model of extremely stressed parts of the concrete structure (a part of the macro-model) .

1 .

Introduction

The nuclear power plant Dukovany (EDU) has been in use under reliable operation for more than 15 years .Within the programme Harmonisation whose aim is to ensure high-quality and safe operation of thi snuclear power plant at least until the year 2025 so for this reason there has been disposed a great number o ftasks concerning various areas. Actual static reliability of concrete structures is besides others one of theproblem of the power plant building part . Considering the fact that concrete structures have to befunctional, safe and reliable for substantial time period after the operation of the nuclear power plant ther ehave been worked out large-scale procedures and models for evaluation of particular - for the nuclea rpower plant reliability - dominantly important structural parts . The problems of evaluation of the concretestructure reliability are solved from the experimental and theoretical point of view .

55

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2 .

Experimental part

Concretes of the load-bearing structures of the primary circuit are effected not only by the mechanica lstress but also by the moisture stress . Moreover, these concrete structures are also subjected to long-periodinfluence of high temperatures during their lifetime . Due to the present temperature and moisture stresses ,new crystalline formations inside the concrete structure can develop (e .g . 11 A-tobermorit) and so one partof evaluation works was focused on observations of the actual physically- mechanical characteristics o fconcrete . The experimental part of the work was especially concentrated on the following areas :

- Determination of the distribution of temperature field on the surface of the concrete structures whic hserves as the boundary conditions for temperature field (calculated by the means of mathematicalmodels )

- Determination of the distribution of moisture field inside the concrete structures . It is used as theconstraint condition for moisture field determination by the means of mathematical mode l

- Determination of the actual physically-mechanical characteristic s- Determination of different degradation effects on the concrete structur e

of the primary circuit .

2.1 . Measurements in situ and laboratory test s

There were carried out

Measurements of concrete moisture (by gravimetric and by neutron method) ,Determination of the physically mechanical characteristics of concrete by laboratory tests o nsamples obtained by drilling (volume mass, modulus of elasticity, stress-strain diagram underthe temperature stress, characteristics of temperature and moisture expansivity, determinatio nof the content of soluble boron salt in concrete etc .) ,Evaluation of radionuclide level activity of concrete ,Measurements of thickness within the non-hermetic internal protective lining in the zones o fthe moisture content measurement of concrete by the neutron method .

Concrete samples were drilled from predefined areas and according agreed schedule within the period1997-2001 (diameter of the samples was 100 mm) . The other samples were fragments of the externalsurface of concrete under the protective lining used for determination of the concrete structure moisture .

The range of measured moisture values within the observed period was roughly the same . However,significant differences were among the moisture values measured on the same areas . This provedconsiderable changes of the moisture conditions - Fig . 1 . It is possible to state that the moisture migrationinside concrete occurs during the time period . From more measurements carried out in the identical area swithin one shutdown it is evident that the moisture varies in time dependency .

56


Following physically mechanical concrete characteristics of the primary circuit were found at the analyse s

- Volume mass (density, thickness) under natural conditions p 2192 to 2330 kg/m3- Compression strength f c 48 .1 to 60.8 MPa,- Modulus of elasticity Ec 25 .2 to 33 .4 GPa .

Furthermore, the temperature and moisture expansivity characteristics were measured and the stress-strai ndiagrams of concrete were verified (including the decreasing branch) under the temperature stress .

Considering the fact that there were prescriptions and records available concerning the evident concretetests within the time of the nuclear power plant construction, it was possible to compare the originphysically mechanical characteristics with the ones after 15 years operation . No significant difference swere found .

2.2 . Conclusions of the experimental par t

From the experimental tests of the steel and concrete samples and from the measurement in situ i tis evident that

- temperatures of the concrete structure exceed 100o C in some areas ,- the migration of moisture inside concrete demonstrates itself within the time of shut-down, which was

found out by comparison of the moisture in the identical areas of the RC load-bearing structure at th eblock 2 ,

- amount of boron was found in concrete by physically chemical tests ,- in the course of the nuclear power plant operation there have not appeared any substantial physically -

mechanical characteristic changes of concrete within the observed structures of the primary circuit ,- the incidence of pitted corrosion was found in the samples of non-hermetic protective lining (it doe s

not have any static importance but it only forms the protection of concrete against the contaminatio nat purification during the shut-down),

- the appearance of artificial radio-nuclides in the samples of concrete taken from the structure of th eprimary circuit was found . The radionuclide content in samples is so low that it does not influence thephysically mechanical properties . Also from the point of view of the State Supervision of Nuclea rSafety (SÙJB) classification there is not dealt with any emitter,

- the appearance of CSH-gels and 11-A-tobermorit in the samples of concrete that only demonstrate sthe present existence of the increased moisture and temperature at the time of its origin .

3 .

Modelling of the structure behaviou r

Consistent with the observation of the structure behaviour of all four blocks of th enuclear power plant it was stated that

- all EDU blocks can be considered to be identically thermally stressed and that is why it is no tnecessary to distinguish the temperature stresses of the individual blocks . The differences oftemperature about ±5 C in the extremely stressed areas cannot be considered as substantial ,

- in some areas of the concrete structure the temperature of 50 oC is exceeded in long terms. It is thetemperature that causes the decrease of the concrete strength in accordance with the standard • SN 731201-86 ,

- locally variable moisture was found in the concrete structures close to the reactor .

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3.1. Types of loading

3.1 .1 . Radiation loading

The detailed analysis of the radiation influence on the concrete structures reliability was carried out -detailed theoretical summary can be found in [4], [5] . According to the American standard ANSI/ANS-6 .4-1985 [6] the radiation influence on the primary shielding is minimal on the condition that the density of th etemperature flow of energy does not exceed 10 10 MeV . cm-2 . s-1= 16 W m2. According to our carried outcalculations, the density of the temperature flow of energy at the primary concrete shielding entry is 71 4times lower than the value considered by the USA standard. The literature states that the radiation flow o fone mW cm-2 results into the temperature increase of concrete by approximately 1 .5 C. On this assumption(and at the validity of the linear dependence between the temperature and energy of the falling dow nradiation), the temperature increase on the surface of the primary concrete shielding of the reactor wouldbe 0.0036 o C .

On the base of these calculations it is possible to neglect the influence of the internal sources of thetemperature inside concrete initiated by the radiation on the temperature field values when it is solved b ythe FEM method.

3.1 .2. Moisture loading

Considering the fact of random moisture migration inside the concrete structure, the influence of th emoisture expansivity on the state of stress of the structure was neglected at simplifying modelling . Uptoday, there have not been available reliable time-dependent changes of the moisture during the tests .

3.1 .3 . Other types of loading

The limiting importance for the state of stress evaluation, function reliability and durability of the concret estructure has the distribution the temperature field inside the concrete structure and by this generated th estate of stress (of course, in the co-action with the mechanical, radiation and moisture stress) . That is whywe solved the problem s

n relating with the specification of a mathematical model of the parts of the primary circuit structures ,which are based on :1. definition of the boundary conditions of temperature field within the mathematical model of th e

primary circuit structures ,2. theoretical analysis of the determination of the concrete structure referential temperature . The aim

was to formulate the theory needed and to define the referential temperature field within th esolved structures in dependence on their thickness ,

3. final solution of the problem considering the influence of random temperature sources insid econcrete initiated by absorption of neutron radiation and by gamma emission ,

4. complementation of some missing data relating to geometry, material characteristics and thecomposition of some structural parts .

n application of the specified model for solution of the particular situation defined after the discussio nwith the nuclear power plant workers ,

n evaluation of the actual static reliability of the concrete structures within the above stated situations .

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3.2 . Model of the StructureBased on the experimentally obtained data (the boundary conditions of the temperature field), followin gamendments on the mathematical model of the primary circuit structures were carried out taking intoaccount the comments issuing from the discussion with some EDU department s

- Geometry improvements of the modelled structure (more exact model) .- Physically-mechanical characteristics improvements of some materials in accordance with results o f

the carried out experimental tests .- Specification of the boundary conditions of the temperature field according to the measured result s

(non-contact thermometers, thermo-vision, standard and non-standard measurements) .-Following calculations were carried out :- calculation of the temperature field distribution ,- calculation of the state of stress generated by the temperature field in the steady state,- calculation of the mechanical state of stress under the static load ,- calculation of the total state of stress (under the temperature and static load) .-Following design states were taken into account :n STANDARD : the standard operating situation that corresponds to the steady (time-independent )

behaviour of the concrete structures during testing procedures when the influence of the reactorstarting-up operation is not substantial (stationary problem). This situation is defined by

the boundary conditions of the temperatures measured (non-contact thermo-meters ,thermo-vision, standard and non-standard measurements performed on the modelled part of thestructure),the dead load, the technology load (machinery parts loading the structures) and by the boro ncontent of a reservoir,consideration of the creep influence on the state of stress generated by the temperature stress (thetemperature loading is assumed as the long-terms one) .LPT 30: the hypothetical operating situation that corresponds to the steady behaviour of theconcrete structures during the testing procedures and that is defined bythe dead load, the technology load (machinery parts loading the structure) and by the boro ncontent of a reservoir,final definition of the temperature boundary conditions that correspond to the temperature fieldincreasing values on the measured ones . There is assumed : the temperature increase in the reacto rzone by 300 C, zero temperature increase on the boundary of the solved situation (solved part)and the linear course of the temperature increment among the boundary values defined by abovementioned two items ,the consideration of the creep influence on the state of stress generated by the temperatur eloading .KPT 30 : the hypothetical operating situation that corresponds to the steady behaviour of th econcrete structures during the testing procedures and that is defined by :the dead load, the technology load (machinery parts loading the structure and by the boro ncontent of a reservoir,final definition of the temperature boundary conditions that correspond to the temperature fieldincreasing values on the measured ones. It is assumed: the temperature increase in the reacto rzone by 300 C, zero temperature increase on the boundary of the solved part and the quadrati capproximation course of the temperature fields of the concrete structure s

--

-

-

-

-

-

-

-

-

-

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n KPT 30 KR : the hypothetical operating situation that corresponds to the steady of behaviour of th econcrete structures during the testing procedures and that is defined as the same as the situatio nKPT 30 with the only difference when the short-time temperature stress is assumed (e .i . the creepinfluence on the state of stress generated by the temperature stress is neglected) .

n LPT 30 KR: the hypothetical operating situation similar to LPT 30 . The short-time temperatureloading is assumed .For details see [1], [2], [3] .

3.3 . Conclusions of the experimental part

From the works that were carried out in the years 1998-2001, which dealt with the problems of th econcrete structures reliability on the primary circuit, can be concluded :

according to the carried out evaluation of the static reliability in accordance with the Czech standard • SN731201-86 at the steady temperature operating regime can be stated that all the structural parts meet therequirements of the calculated stresses for following combinations :

- situation STANDARD (dead load + technology + temperature in the steady operating regime )- dead load + technology + temperature in the steady operating regime increased by 30 0 C in the

reactor zone (linear course, loading situation LPT 30) ,- dead load + technology + temperature in the steady operating regime increased by 30 0 C in the

reactor zone (quadratic course, loading situation KPT 30) .n satisfactory compliance among the results of the numerical temperature field solution an d

measured values was found .

4 .

Conclusion

By solving of the described interactive problem we ca nâ obtain values of the deformations and internal forces of the concrete structur eâ analyse the influence o f

• the radiation, temperature and static stress acting on the structure• the physical non-linear behaviour of concrete• the shrinkage and creep of concrete .

The results of analyses can serve as the base for correct computing of time reliability and lifetimeprognosis of the concrete load bearing structures .

It was confirmed that the set-up mathematical models would be possible simply to extend for the casesusing the simulation method and for evaluation of the concrete structure reliability in which some of theinput data are considered as random . At present, input statistic data are gathered for application of thesemodels . But the substantial complication will be – regarding the extent of the solved task - the tim edemand of the calculation, but even this problem can be solved – [7], [8] . Application of the non-linearmodelling is another sphere of the result accuracy specification .

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In regard to the extreme loading (mechanical, temperature, moisture and radiation), owing to importance o fthe concrete structures and in view of extremely high demands on the reliability (that must be regularl ycontrolled during the nuclear power plant operation and even after its closing as well), there appears as th eonly possibility of the experimental monitoring combination of important properties of a concrete structurewith numerical verifying of the structure actual reliability by the help of mathematical modelling (withregard to the actual physically-mechanical characteristics) in the future . In connection with this method o fthe reliability conclusive evidence there are being prepared :

- application of the totally reliable access for the reliability evaluation of the concrete structure s(mathematical modelling, simulation methods – input data considered as random values/fiel d

- preparation and completion of a complex of tests for fast models of ageing (degradation processes ,complex of referential samples) .

Acknowledgements

This contribution has been prepared on the base of the scientific research order „Problems of ThermallyStressed Structures in the Nuclear Power Plant in Dukovany“ and the research project CEZ 322/98 2 6100007 „Theory, Reliability and Defects of Statistically and Dynamically Stressed Structures“, Faculty o fCivil Engineering.

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Literature

1 ] Analysis of problem of loading of building structures in power plant with reactor VVER 440 . Final report 1997,ÙBaZK FAST VUT v Brn , 12/1997

2 ] - [3] Temperature and mechanical stress and reliability of concrete structures . Final reports 1998, 1999 an d2000. ÙBaZK FAST VUT v Brn , 12/199 8

[4] Kaplan M.F . : Concrete Radiation Shielding (Nuclear Physics, Concrete Properties, Design and Construction) ,Longman Scientific & Technical, England, 198 9

[5] Hubbell J .H . : Photon Mass Attenuation and Energy-absorption Coefficients from 1 keV to 20 MeV, NationalBureau of Standards, Washington, USA, 198 1

[6] ANSI/ ANS – 6 .4 – 1985, American National Standard for Guidelines on the Nuclear Analysis and Design ofConcrete Radiation Shielding for Nuclear Power Plants, American Nuclear Society, Illinois, USA, 198 5

[7] IAEA: Safety Aspects of Nuclear Power Plant Ageing . IAEA-TECDOC-540, 199 0[8] IAEA: Data Collection and Record Keeping for the Management of Nuclear Power Plant Ageing . Safety Serie s

No. 50-p-3, 1991

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Fig. 2 : Distribution of the temperature field on the concrete structures surfac e

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Fig. 3 : State of stress (principal stress) of the concrete cantilevers initiated by the temperatur eloading

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Fig. 4 : Total state of stress (principal stress) of the concrete cantilevers (interaction of mechanical an dtemperature loading

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Efficient management of inspection and monitoring data for a better maintenance of infrastructur e

Marcel de Wit 1 Gilles Hovhanessian2

Area Manager

Deputy general manager ,

Advitam, Northern Europe

Advitam, Pari s

Keywords : management, monitoring, inspection, maintenanc e

Abstract

In North America, Europe and Japan, government agencies and larg eprivate owners are now facing the challenge of maintaining, with limitedresources, large stocks of vital structures like traditional and nuclea rpower plants, Cooling towers but also highways, railways, bridges ,dams, harbors, industrial facilities etc… These structures arerepresenting a large amount of money, have not been designed to b eeasily repaired or replaced, and are getting older and more vulnerable .

People involved in structure management have developed extensiv etechnical methods and tools to monitor the condition of the structure and establish the diagnosis . Eachauthority has been developing is own inspection maintenance procedures, taking into account thei rspecificity, their different priorities, safety requirements, resources and range of competence .

In most cases visual inspections are used to detect deteriorations, to rank structures, define priorities ,estimate repair costs, etc… These visual inspections require to record, report, analyze and store for year slarge quantities of data (inspection records, drawings, photos…) and it is easy to get lost in the clerica lwork. Moreover a number of decision steps (inspection record, ranking of defects, long-term analysis) ar estill highly subjective and can greatly affect the quality of the final diagnosis .

An inspection-based management software system has been developed to optimize this process an dprovide decision-makers with objective information on the condition of the infrastructure . The system is acomprehensive management system which integrates : database of structural defects, on-site computerize drecord, analysis, maintenance, diagnosis, repair and budgetary functionalities .

This paper describes the basic functions and benefits of the system .

1 [email protected] ghovhanessian@advitam-group .com

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1 . LIMITATIONS OF CONVENTIONAL INFRASTRUCTURE MANAGEMENT PROCESS

The characteristics and limitations of, still widely used conventional structure management process arelisted below:

- Design data (drawings), inspection data, detailed investigation data and repair data are not storedin a single system.

- Inspection frequency for a given structure is based on the type and age of the structure . It is rarethat the date of inspection is based on the results of the previous inspection .

- Before inspection, inspectors must prepare inspection drawings - very often original desig ndrawings are not available and inspectors must spend time to make new drawings .

- During inspection : the inspectors take hand-written notes of the defects . Inspectors usually do notbring with them the heavy reference manuals .

- Back in office the inspectors copy the deterioration onto the structural drawings, along with theirdimensions and characteristics . Sometimes these data are stored electronically (excel sheets an dCAD drawings) .

- In accordance with the inspection manual, a ranking indicator or a comment is affected to eachdeterioration . The inspectors then establish reports that are transmitted to the engineers in charg eof the analysis.

- The engineers receive several reports from different inspectors . They may have difficulties withinconsistent data, inhomogeneous ranking systems, unreadable handwriting or confusin gdimensions . However on the basis of these reports the engineers must estimate the condition of thestructures and recommend actions for maintenance or repairs .

- When needed, detailed investigations are performed by specialized consultants and specialize dcontractors propose repair solutions .

- Maintenance and repair costs are then presented to decision makers .

After reviewing the above points, it becomes clear that even with a clear inspection manual and anefficient organization, conventional infrastructure management process allows too much room fo rsubjectivity and conventional infrastructure management is expensive .

2. OBJECTIVES OF THE INSPECTION-BASED MANAGEMENT SOFTWAR E

The inspection-based management software has been developed with the following objectives :- improve the overall efficiency of the maintenance process ,- reduce the cost of maintenance process at all steps ,- build a comprehensive database system which integrates all steps of the maintenance proces s

(inspection preparation - inspection - reporting - analysis - repair - budget) ,- facilitate the task of inspectors ,- assist engineers in the compilation and analysis of large quantities of data ,- allow easy access to all data at any step of the engineering and decision process ,- provide decision makers with valuable and objective information on the condition of th e

infrastructure, on which they can base and justify their decisions .

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3. BASIC DESCRIPTION OF THE INSPECTION-BASED MANAGEMENT SOFTWARE

The inspection-based management software consists of several components specifically designed t ohandle the tasks of each people involved in the maintenance process (Table 1) .

Software components Tasks Designed fo r

Infrastructure Management

- database of detailed informatio non structures (drawings, design-construction-inspection-repairdata) ,- ranking of structures accordingto preset rules ,- management of the database,

- operators of structure ,- consultants ,- specialized inspection companies

- budgetary tools ,- scheduling of tasks ,- management of repair works

Inspection Management- preparation of inspections ,- transfer of data from mainframeto mobile inspection units

- inspectors ,- specialized inspection companies ,- consultants

Inspection- on site recording of deterioratio n(on pen-touch light computers)reference availabl e

Photo-based inspection- time-effective survey ofdeterioration based on photos

Report- reporting of site-records ,- automated standard report

Analysis- advanced analysis functions ,- detailed investigation ,- detailed repair definition

- engineers ,- consultants ,- contractors

Table 1 : Components of the inspection-based management software

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-

--

-


Fig.1 : General information tab

3.1 Infrastructure management software

With the infrastructure management software, operators can organize their inventory of structures . Thesystem allows to build a database including :

general information about each structure (design, construction, location, pictures, drawings – se eFig .1) ,detailed check-lists for each structural component (see on Fig .2 an example for expansion joints) ,damage criteria for ranking of deteriorations (see Fig .2 where limit values for joint opening aredefined) ,catalogue of repair solutions and corresponding costs and durations (see Fig .3) .

The software has been built so that all parameters can be changed and adjusted to the specific usage o feach industry/administration/operator. For example : the limit criteria for crack opening is smaller in nuclea rcontainment vessels than in highway tunnels. Such a limit criteria can be set in accordance with th ecorresponding regulations .

.

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Fig .2 : Definition of check-points and damage criteria for expansion joint s

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Fig.3 : definition of tasks, with corresponding costs and scheduling

3.2 Inspection softwar e

The inspection software is designed to simplify the task of inspectors and ensure coherent and regula rdeterioration surveys and maintenance records (Fig .4) . Running on pen-touch computers (Fig .5), theinspector can access at any time :

- the inspection reference manual ,- the drawing of the inspected structure ,- the history of the deterioration that was recorded in past inspections ,- the maintenance check-list for each structural component .

When the inspector detects a deterioration, he draws with the pen directly on the computer screen the shap eof the deterioration. He can use a reference of more than 200 deterioration types classified in families . Inaccordance with the inspection manual, the software then requests the inspector to measure and record acertain number of parameters to describe the deterioration (dimensions, color, humidity level…) . He mayalso want to take pictures of the deterioration . The software saves all this information in the database :

- the deterioration type ,- its graphical representation in the CAD drawing,- its dimensions and other specific parameters ,- any picture of the deterioration.

Because these data are linked together in the database, it will be easy to access and sort such information a tany step of the maintenance process, even years after the inspection .

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Fig .4: Site inspection software running on pen-touch computers

3.3 Photo-based inspection software

In some cases when access is difficult, survey of deterioration can be done using high quality digita lpictures . The photo-based inspection software has the following functions :

- correction of the lens deformation of the picture,- deformation/scaling of image,- on-scale insertion of image data into CAD drawing ,- highlighting of typical deteriorations (rebar, cracks… )

Photo-based inspection is often a cost-efficient alternative for the inspection of large structures (dams ,cooling towers, etc… )

Fig.5: Site inspection using pen-touc hcomputers

Fig.6: Photo-based deterioration survey

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3.4 Reporting software

Immediately after an inspection has been performed, the inspectors can edit standard reports with thereporting software :

- output of drawings with deteriorations (Fig .7) ,- tables ,- pictures .-The process is automated and the output format can be adapted .

3.5 Analysis software

Designed for engineers in charge of the analysis of data, the analysis software consists of a set of tools for :- browsing and sorting of data,- evolution of one deterioration or a group of deteriorations ,- comparison of similar structural components ,- evolution of the condition of one structure, evolution of the stock of structures ,- identification of deterioration to be repaired in priority ,- bill of quantities for repairs ,- assistance in the diagnosis .

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4. BENEFITS AND DIFFICULTIE S

The direct benefits are cost reduction and improved efficiency due to :

- organized database of consistent data (structural, inspection, maintenance and repair data) ,

- easier access and sharing of information ,

- time savings for preparation of drawings, reporting and analysis of data ,

- long-term management.

Whereas the main difficulties in implementing the system are :

- staff need to be computer literate ,

- drawings that exists only on paper must be digitalized (scanner) or redrawn with CAD ,

- older data should be input in the new system .

The system is therefore easier to implement on recent structures because CAD drawings are available andolder data is smaller : the system has been used from the beginning on the Tagus estuary crossing inPortugal (Vasco de Gama bridge) .

However, the extra work required to input paper-based drawings and older data can be recovered throug hcost and time savings at all steps of the process . During the inspection and analysis of the Zilwaukee bridg e(a twin 2 .5 km-long precast segmental viaduct carrying I-75 over the Saginaw River, Michigan, USA), th esystem proved to be very cost-effective. Every day the inspectors sent the data by email for review by th eProject Manager and the analysis team, located 3,000km from the bridge . Data from previous inspection swas later inputted electronically so that it can be compared, sorted and visually displayed along with ne wdata.

5. CONCLUSION

Structure management is an increasingly important concept for structure owning authorities or privat ecompanies around the world today . Each owner has been developing its specific health condition indicatorsystem, allowing to express the structural condition of a structure by some quantitative measure, to monito rdurability, safety and to decide at what point action needs to be taken .

A management software that would integrate all the steps of this maintenance process can dramaticall yoptimize its efficiency through easier management, storage, sharing and analysis of the structura linformation . While such a concept is not new, this paper presented in detail the capabilities and benefits a sthey were observed in actual large scale implementation of the inspection-based management software .

REFERENCES :

Stubler, Domage, Youdan, "New Developments in Structural Monitoring and Management for Bridges" ,IABSE Conference, Cairo .

Stubler, Le Diouron, Elliott, "New Tools to Listen and Watch Structures for a Complete Monitoring" ,Proceedings of The Korea Institute for Structural Maintenance Inspection, Vol .5, No .1, May 2001 .

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AGING PROCESS OF A GOOD CONCRETE DURING FORTY YEAR S

Dr. Peter Lenkei

Pécs University, College of Engineering (Hungary)

A prestressed concrete truss (Fig . 1) was used to cover an uranium ore processing (concentrating) hall in anuclear industry plant . The environment of the hall was slightly aggressive, containing sulphuric acid an dcarbon dioxide. Due to the uranium oxid dust all the equipment and the reinforced concrete structures wer ecovered with a special paint coating for easier decontamination .

After 40 years of service the uranium ore mining was terminated and the hall was demolished (Fig . 2) .

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During the life span of the structures several investigations were made . Destructive tests (DT) were madeafter construction in 1960 (200*200*200 cubes), in 1990 and in 2000 thorough non-destructive test s(NDT) were carried out. Finally, after demolition another DT was made, core samples were taken from thetruss. The results converted to 150*300 cylinder strength are shown on Fig . 3 .

Likewise after the demolition parts of the prestressed bottom chord of the truss were surveyed and neitheron the prestressed wires, nor on the other reinforcement traces of corrosion could be find (Fig . 4).

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Fig. 4 Surface of prestressing wires and rebars

The results demonstrated, that the concrete aging process was characterized by a definite increase over theinitial strength and even after 40 years the concrete strength was 3% higher of the initial strength . Mostprobable the NDT gave a little higher results over the true strength values . Neither visible cracks orcarbonization, nor reinforcement corrosion were detected .

CONCLUSIONS

1. Initially good quality concrete and reliable concrete cover, with sufficiently maintained paintcoating could guarantee the long term life span of prestressed concrete structures .

2. Even the correct NDT may slightly overestimate the concrete properties .

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SESSION B: STATE OF THE ART &FUTURE DEVELOPMENTS (Continued )Chairman: Mr. Jean-Pierre Touret, EdF, (France)

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THE USE OF ACOUSTIC MONITORING TO MANAGE CONCRETE STRUCTURES IN THENUCLEAR INDUSTRY

Marcel de Wit, Gilles Hovhanessia nAdvitam

ABSTRACT

Concrete and steel are widely used in containment vessels within the nuclear industry . Both are excellentacoustic transmitters . In many structures tensioned wire elements are used within containment structures .However, tensioned wire can be vulnerable to corrosion . To reduce the probability of corrosionsophisticated protection systems are used . To confirm that the design strength is available through time ,extensive inspection and maintenance regimes are implemented .

These regimes include tests to confirm the condition of the post-tensioning, and pressure tests (leak tests )to verify the performance of vessel .

This paper presents an acoustic monitoring technology which uses widely distributed sensors to detect an dlocate wire failures using the energy released at failure . The technology has been used on a range o fstructures including post-tensioned concrete bridges, suspension bridges, buildings, precast concret ecylinder pipelines (PCCP) and prestressed concrete containment vessels (PCCV), where it has increase dconfidence in structures and reduced maintenance costs .

Where the level of ambient noise is low then SoundPrint® acoustic monitoring can detect concretecracking . This has been shown in PCCP pipelines, on laboratory test structures and also in nuclea rstructures. The programme has shown that distributed sensors can locate internal cracking well befor ethere is any external evidence .

Several projects have been completed on nuclear vessels . The first has been completed on an Electricité deFrance (EDF) concrete test pressure vessel at Civaux in France . The second at the Sandia PCCV TestVessel in Albuquerque, New Mexico, USA, which involved the testing of a steel lined concrete vessel .The third was on a PCCV in Maryland, USA .

Acoustic monitoring is also able to monitor the deterioration of post-tensioned concrete structures as aresult of seismic activity . Summary details of a case history are presented .

1. INTRODUCTION

Continuous acoustic monitoring has been used since 1994 to monitor failures in bonded and unbonde dtendons in post-tensioned structures, where it has shown major benefits in confirming the performance o fstructures, increasing Client confidence and reducing maintenance costs . To extend the application of thi stechnology to the monitoring of concrete cracking required that the effectiveness of the principles an dmethods was evaluated for each structural type .

For acoustic monitoring technology to function in a particular environment it must be shown that th esignals generated by cracking can be detected above general noise levels and distinguished from event swhich are not of interest . Furthermore, to assess the structural implication of each event it is generall yimportant to be able to locate the source of each emission . Provided with high quality data of this type, th eengineer can appraise a structure with knowledge of the actual failures in damaged elements, and their

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location, in the entire structure over the monitoring period . The alternative, to base the assessment on aphysical inspection at a sample of locations, leads to uncertainty when for practical and economic reason sthe number of inspection points is limited. Monitoring the entire structure may also reveal failures no tdetectable by a conventional investigation .

In many applications the acoustic data is transmitted over the Internet for processing and analysis . Afterprocessing and quality control checks, the data can be made available on a secure section of th eSoundPrint® website, allowing owners rapid independent access to their database of results .

SoundPrint® acoustic monitoring systems have also been placed on structures, which are in active seismi czones. Rapid status reports on wire failures / structural damage allows Owners and Regulators to assess th econdition of a structure within a few hours .

The technology is useful in providing cost-effective long-term surveillance of both unbonded and groute dpost-tensioned containment structures. This paper shows how the technology can also monitor thecracking of concrete structures which are subject to low levels of ambient noise .

2. DEVELOPMENT OF CONTINUOUS ACOUSTIC MONITORIN G

The principle of examining acoustic emissions to identify change in the condition of the structural element sis not new. However, until recently, continuous, unattended, remote monitoring of large structures was notpractical or cost-effective. The availability of low-cost data acquisition and computing hardware,combined with powerful analytical and data management software, resulted in the development of acontinuous acoustic monitoring system called soundprint®, which has been successfully applied tounbonded post-tensioned structures in North America since 1994 .

Corrosion of the steel strands in these post-tensioned structures has become a concern for designers andowners . As with grouted post-tensioned bridges, the extent of corrosion is not known, primarily because o fthe difficulty of identifying corrosion due to the inaccessibility of the corrosion sites, the lack of externa levidence and the limited spatial coverage of intrusive inspections .

The SoundPrint7 system uses the distinctive acoustic characteristics of wire breaks to separate them fro mother acoustic activity on a structure . Using a combination of instrumentation, data acquisition and datamanagement, it is possible to identify events, as well to locate the failure and time of failure .

This concept allows the non destructive identification of broken strands, so that these strands can bereplaced periodically as part of a long term cost effective structural health programme . In addition, anunderstanding of the condition of the steel wire elements allows the life of the structure to be extended .

A typical system includes an array of sensors (Figure 1) connected to an acquisition system with coaxia lcommunication cable . The sensors are broadband piezo-electric accelerometers fixed directly to th econcrete slab . Sensor locations are chosen so that an event occurring anywhere on the slab can be detecte dby at least four sensors . Sensor spacings range from 1 per 60 square meters for fully grouted slabs up to 1per 100 square meters for ungrouted tendons. Multiplexing techniques are able to acquire data from man yhundreds of channels on 32 acquisition channels .

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Time

Frequency

Figure 1 - Standard sensor for buildings, bridges and Figure 2 - Time domain and frequency spectrum plot sparking structures

of wire break detected by sensor 10 .0 m. from event

Figure 3 - Time domain plot showing relative arrival Figure 3 - Time domain plot showing relative arriva ltime of signal at different sensors

time of signal at different sensor s

Using several characteristics of the acoustic events including frequency spectrum it is possible to classif ywire breaks and to reject environmental noise .

By analyzing the time taken by the energy wave caused by the break as it travels through the concrete t oarrive at different sensors, the software is able to calculate the location of the wire break, usually to withi n300 - 600 mm of the actual location . Independent testing showed the system to be 100% correct whe nspontaneous events classified as "probable wire breaks" were investigated . Figure 2 shows a typica lacoustic response to an unbonded wire break at a sensor 10 .0 m from the break location . Figures 3 and 4illustrate how the system locates events .

SoundPrint® site systems download all data automatically using the Internet to the Calgary processingcenter . This allows the cost of data transfer to be minimized . All data can be viewed by the owners teamdirectly on the Pure Technologies secure web site . This allows the owner to review areas of concern in

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parallel with the generation of routine reports . Various levels of alarms can be triggered semi-automatically using e-mail, automatically voice activated phone alarms, etc .

Presently, over 300,000 square meters of unbonded post-tensioned slab in twenty structures, five bridgesand almost 100km of large diameter water pipe are being simultaneously monitored . The analyticalsoftware is capable of automatically generating reports summarizing the time and location of wire break sand other significant events . The operating efficiency of the system over the monitoring period is alsorecorded.

3. MONITORING OF WIRE BREAKS IN GROUTED POST-TENSIONED BRIDGE S

Acoustic monitoring has been used in a wide range of applications including suspension and cable staybridges (reference J .F . Elliott), and pipelines (reference Mark Holley) .

The technology has also been applied on many post-tensioned concrete bridges as described at thi sconference (reference Carlyle, Adkins, Youdan) .

4. MONITORING OF CRACKING DEVELOPMENT IN CONCRETE STRUCTURE S

Description of Concrete Projects

During the UK TRL grouted post-tensioned bridge evaluation program, the developers of the acousticmonitoring system, Pure Technologies Ltd (Pure) and TRL had the opportunity to evaluate the applicatio nof the method to crack development in a partially hollow reinforced concrete beam specimen. Thisspecimen was tested with three-point loading .

Dr. Walter Dilger of the University of Calgary provided access to a flat post-tensioned slab specimen . Theslab was 5 m by 10 m by 150 mm thick supported by 3 columns and tested in shear .

Electricité de France allowed access to a large-scale model prestressed concrete containment vessel a tMaeva being tested with internal pressure . This vessel has part of the perimeter wall instrumented fo rcrack detection. This specimen had previously been tested to the same pressures used in the curren texperiment.

At the Sandia National Laboratories in Albuquerque, New Mexico, a ¼ scale test vessel was pressured t othe full ‘failure’ test pressure in a modification of a standard leak rate test . The objective was to monito rconcrete cracking, tearing of the liner, and gas leakage .

Finally in Maryland, US an operating PCCV was tested during an Integrated Leak Rate Test to determin eif any wire failures were recorded .

Transport Research Laboratory, UK

A partially-voided reinforced concrete bridge beam was loaded as shown in Figure 5 .

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Microphones / accelerometers were installed at six locations on a voided beam specimen . Stress wasapplied as three point loading . Emissions were noted at all loads and continued throughout the test .

Figure 5 - Arrangement of Test Specimen VS1 7

Figure 6 – Location of Events from 0 kN to 30 kN (0 to 6,700 lbf)

Figure 7 – Location of Events from 30 kN to 100 kN (6,700 lbf to 22,500 lbf)

Figure 8 – Location of Events from 100 kN to 200 kN (22,500 lbf to 45,000 lbf)

Result s

As used here, ‘cracking noise’ means the generation of acoustic events associated with the propagation ofcracks, some of which were not visible . Amplification through the data acquisition system producedaudible cracking noise throughout most of the test . No sounds were heard or recorded during periods whe ndisplacement had stopped . Acoustic events were located as plotted in Figures 6, 7 and 8 .

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It was noted several times during the test that the locations of cracks could be determined from the acousti cdata before the cracks became visible . On most occasions, the operators of the acoustic equipment wer eable to direct researchers to the area where cracks had occurred, resulting in the visual confirmation ofcracks at those locations .

Crack Monitoring at University of Calgar y

Procedure

Ten accelerometers were attached to the underside of the test slab . The slab-column arrangement is shownin Figure 9 and sensor locations are shown in Figure 10 . Lateral motion of the slab was commenced andthe resulting cracking events were heard and recorded . Upon first loading of the specimen, a very largenumber of small emissions were heard . This is known as the Kaiser . This effect describes the generatio nof acoustic events coincident with initial load sharing and redistribution when a concrete specimen is firstloaded to a given level. Subsequent unloading and reloading to the same level will not produce ne wacoustic events until the previous maximum load is exceeded. The rate of occurrence of these emissions i sestimated at between 10 and 100 per second in the specimens tested at the rate of loading used . A sampleof the time-domain data is shown in Figure 11 . Each graph represents the output of one sensor.

10 m

Direction of Load Applicatio n

Figure 9 - Slab Arrangement – Specimen UCS 1

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Figure 10 – Sensor Locations Figure 11 – Time Domain Graphs

Result s

Amplification through the data acquisition system produced audible cracking noise throughout most of th etest . No sounds were heard or recorded during periods when displacement had stopped . The locations ofthese events are shown in Figure 12 .

As was the case with the TRL test, it was possible to direct researchers to the location of cracking befor ethe cracks were visible . The locations of cracks identified by the acoustic system coincided with th econfirmed locations as determined by the researchers .

Figure 12 – Event Location sMaeva Model Containment Vesse l

The Maeva vessel was built for other purposes relating to internal pressure testing . The vessel consists of acylindrical wall with an internal diameter of 16 .0 m (52 .5 ft .) and an external diameter of 18 .4 m (60 .4 ft .) .The floor and roof of the vessel consist of concrete slabs connected by four columns each containin gsixteen x 75 mm (3 in .) high-strength steel Macalloy bars . The concrete wall is enclosed by a watertigh tsteel bulkhead. Instrumentation has been installed on two panels of the vessel to confirm the ability of th eacoustic monitoring system to detect cracking of the concrete as pressures change . The programmeincludes a medium pressure test to 5 .66 bars (82.1 psi) (which has been completed) and a high-pressure testto 10 bars (145 psi) (which is to be completed later) . Because the vessel had previously been pressurize dto a greater pressure than was used in the medium pressure test, the vessel was not expected to produce theKaiser effect cracking .

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Sixteen sensors were attached to two contiguous wall sections of the vessel in the pattern shown in Figure13 . The sensors were attached to the outer perimeter of the vessel in the annular space between the inne rand outer wall . The sensors are connected to a data acquisition unit located in a building 200 m (656 ft . )from the structure .

4 .94-~ . . 4 .56FR1Z1AFR1Z1A01 FR1Z1A0 2

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Figure 13 – Sensor Layout and Location of Acoustic Cracking Events onTest No. 3, June 1999

First Result s

Sixteen acoustic cracking events were detected and located. Of these, events numbered 2, 6, 14 and 1 6occurred outside the area monitored and therefore could not be accurately located. Locations of events areshown on Figure 13 . The largest cracking events were detected at locations 1, 5, 6, 12 and 15 .Comparison with a known impact from a Schmidt hammer suggests that some of the cracking event sreleased approximately 1 Joule of energy. Event #1, the first large event, occurred at an internal pressur eof 3 .01 bars (42.66 psi) . Maximum pressure achieved was approximately 5 .66 bars (82.09 psi) . Events 1 5and 16 occurred after the pressure began decreasing, at 5 .65 bars (81 .95 psi) and 5 .46 bars (79 .19 psi)respectively. Time domain and frequency spectrum plots for Event #1 are shown in Figures 14 and 15.

To confirm the relationship between acoustic events and cracks it is common to correlate the visiblesurface evidence of cracks and the measured locations of the acoustic cracking events . This process i slimited by the fact that cracking events recorded by SoundPrint ® may reflect cracks that are internal to thestructure. Also a visual inspection will normally only record cracks that are of the order of 0 .2 mm (0 .008in.) wide, whereas SoundPrint® will record cracks finer than this . For this test vessel two additional factor sare relevant . The first is that no crack survey was taken before this latest phase of the testing commenced,and the vessel had previously been taken to higher pressures than were used in this phase of testing. Thesecond is that the internal surface of the concrete had been treated with a thin layer of epoxy, which wouldhave the effect of reducing the clarity of the any crack survey . However, a preliminary crack survey wascarried out after the medium pressure tests and this survey will be repeated after the high pressure tests .

The results of the first phase of acoustic monitoring have confirmed that the system is workingsatisfactorily, can identify acoustic cracking activity and is able to locate the source of such crackingevents .

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Figure 14 - Time Domain Plot of Sensors Figure 15 - Frequency Domain Plot of SensorsResponding to Crack #1 .

Responding to Crack #1 .

Outline Details of the Sandia Test

In a US project sponsored by MITI and the National Research Council (NRC), a one-quarter scale mode lof a steel lined containment vessel has been constructed at Sandia National Laboratories in Albuquerque ,New Mexico. The model is approximately 16 .5 m (54 ft .) high and 11 m (36 ft .) in diameter . Theobjective of the work was to validate numerical simulation methods by comparing measured to calculatedresponses well into the inelastic regime, up to and including failure . The intention was to map thedevelopment of strains and eventual damage as the pressure in the vessel is brought above the designpressure of the vessel . Although more than one thousand strain and other gauges were installed on th evessel, much of the vessel was not monitored by strain gauges . Acoustic monitoring was being installed onthe entire vessel wall area as part of the programme to detect tendon failures and with the hope o fmonitoring concrete cracking and liner tearing/leakage . The system has been specially configured tostream data offsite to a back-up computer incase the onsite unit is destroyed during the test . Preliminarydetails of the test are reported by Hessheimer 5 . Results of this test will be published first by the NRC in2001 . See Appendix A for update to June 2001 .

PCCV Maryland

This vessel is approximately 42m diameter reinforced with two sets of circumferential tendons . A total of204 tendons are present ; a normal tendon is comprised of 90No. 6mm wires .

The vessel was monitored with 36 distributed acoustic sensors, which were placed on the external concretesurface . The vessel was tested during an Integrated Leak Rate Test . Surveillance of the vessel was by on-site monitoring in real time during the test . During the ILRT zero wire breaks were recorded. To confirmthe performance of the acoustic monitoring system events were generated using impacts with a simila renergy (and a similar acoustic signature) to a wire break . Impacts made at anchorage cans, were correctlyidentified and located . This was quite an achievement as some of the cans were less than 1m apart. SeeAppendix A for update to June 2001 .

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5. SEISMIC MONITORING

In numerous applications Owners need to know the effect of a seismic event on their structures . This i sparticularly important in seismically active zones and in the nuclear sector where seismic standards arerigorous. In California where there is a statutory requirement to report on the effect of a seismic even twithin 24 hours, SoundPrint ® acoustic monitoring has been used to provide rapid and relevant details to th eOwner .

East Bay Municipal District of Southern California - Brookwood Reservoi r

Above ground prestressed water storage tanks are common in many areas of the world, includin gCalifornia . Failure of these tanks due to corrosion or other factors can be catastrophic . To investigate theusefulness of long term acoustic monitoring as a management tool, the East Bay Municipal District ofSouthern California commissioned the installation of a monitoring system on the Brookwood reservoir, a10 million litre (264,00 US gal) capacity tank in the Walnut Creek area East of San Francisco .

As part of the commissioning process, two individual wires were corroded to failure and the result smonitored with the system installed there . Both wire breaks were detected and located successfully within300 mm (1 ft .) of the actual location.

At 18 :06 on 17 August 1999 there was a Mercalli 5 earthquake in the Bay Area of California . At 23:04SoundPrint® recorded 1 wire break at the water tank resulting from the earthquake some 30-km (19 mi . )distant. Details of this minor damage were e-mailed to the owner within hours . With this data the ownerwas able to report quickly, and positively, recording only minor damage . With the proven success of thesystem the owner is planning to extend the number of structures monitored .

If many sites were monitored, the owner can also use the rapid notification capability of the system todirect emergency repair teams to the areas where most damage has occurred in the event of a larg eearthquake .

6. SUMMARY

Continuous remote acoustic monitoring has been used successfully to determine the time and location o fwire breaks in prestressed structures. Testing of the technique as a method of detecting cracking i nconcrete structures subject to loading has been carried out on post-tensioned and reinforced concret estructures of different configurations . On two of the three structures tested, the locations of cracksidentified by the monitoring system were confirmed by visual inspection . Crack development was detectedby the monitoring system before cracks became visible .

The SoundPrint® acoustic system has been shown to provide reliable continuous remote monitorin gcapability and the ability to determine the times and locations of both prestressing failures, concret ecracking and liner tearing/leakage. These capabilities are useful in providing cost-effective long-ter msurveillance of unbonded and grouted post-tensioned containment structures, and of cracking within theconcrete itself where ambient noise levels are low .

SoundPrint® has been shown to be a valuable management tool where rapid monitoring of seismi cresponse is required .

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ACKNOWLEDGEMENT S

The authors thank TRL and the UK Highways Agency, Dr. Walter Dilger, University of Calgary ,and Electricité de France for use of these data . The authors wish to thank the many individual sand agencies who contributed to the gathering of these data .

REFERENCE S

1. J.F. Elliott . Continuous Acoustic Monitoring of Bridges . International Bridge Conference ,1999, Pittsburgh, Pennsylvania IBC-99, pp. 70 .

2. M.F.Hessheimer, D . W. Pace, E. W. Klamerus, T . Matsumoto, and J . F. Costello .Instrumentation and Testing of a Prestressed Concrete Containment Vessel Model . 14th

International Conference on Structural Mechanics in Reactor Technology, Lyon, France ,1997, pp 792-1, 792-9 .

3. M. Holley, Acoustic Monitoring of Prestressed Concrete Cylinder Pipe . American Society ofCivil Engineers, Pipelines in the Constructed Environment, California, USA, 1998, pg 46 8

4. F. Carlyle, A. Adkins and D.Youdan. The Use of Acoustic Monitoring to Extend the Life o fPost-Tensioned Overbridges at Huntingdon and Mossband, UK . Structural Faults & Repai r2001,9thInternational Conference, July 2001, London .

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Concrete Properties Influenced by Radiation Dose During Reactor Operatio n

Takaaki Konn oTechnical Counselor

Secretariat of Nuclear Safety Commission

ABSTRACT

The radiation dose effects on the physical, chemical and mechanical properties to the biological shieldin gconcrete of the Japan Power Demonstration Reactor (JPDR) were studied to obtain useful information fo rthe plant life management strategy of commercial nuclear power plants . The JPDR was passed 25 year sfrom the construction and performed 957 days operation and the total reactor operating time 14,230 hours .The cumulative radiation dose rate on the biological shielding concrete of the JPDR was estimated a sequivalent with the one that from the current commercial nuclear power plant after operated 40 years . Highradiation dose is a special unique environment for the concrete structures in nuclear facilities . Theevaluation of the radiation dose effect to the concrete structures in the nuclear power plants is importan tfactor for the plant life management strategy considering aging of the concrete properties during plan toperation . Usually, studies on concrete properties influenced by the radiation were performed under th etest condition of short term and high irradiation rate . The test results under the condition of long term andlow irradiation rate for the concrete are rarely exist . This study was conducted using the actual concretesamples from the JPDR biological shielding concrete obtained when the plant was decommissioning . Themaximum fast neutron and gamma ray at the reactor side surface of the biological shielding concrete ar e1 .11 × 10E+18 n/cm2 and 4 .77 × 10E+18 Gy, respectively, at the level of the reactor core . The test resultsshowed that the compressive strength of the concrete samples were not decreased by the radiation exposur ewhich was rather shown the tendency to increase along with the fast neutron fluencies within the test rangeto 10E+17 n/cm2. The test results showed the biological shielding concrete with steel lining have gooddurability in the test range of radiation exposure dose rate in spite of affected with the heat generationwithin the shielding concrete by the neutron and gamma ray flux .

INTRODUCTION

The evaluation of the radiation dose effect to the concrete structures in nuclear power plants is importantfactor for the plant life management strategy considering aging of the concrete mechanical propertie sduring plant operation . Usually, studies on the concrete properties influenced by radiation exposure wereperformed under the accelerated test condition of short term and high irradiation rate . Test results under thecondition of long term and low irradiation rate for the concrete using the actual concrete specimen sextracted from operating nuclear power plants are rarely existing . The study of radiation dose effects on thephysical, mechanical and chemical properties of concrete was performed using the actual biologica lshielding concrete sampled by coring from the Japan Power Demonstration Reactor when the plant wa sdecommissioned . The Japan Power Demonstration Reactor (JPDR) was the first nuclear power generatio nreactor of the rated reactor power 45 MW in Japan that was passed 25 years from the construction at th e1986 and performed 957 days operation that the total reactor operating time was 14,230 hours when theplant was on the decommissioning . The cumulative radiation dose rate on the biological shielding concreteof the JPDR was estimated as equivalent with the one from the current commercial nuclear power plan tafter 40 years operation. High radiation dose is a special unique environment for the concrete structures i nthe nuclear facilities . The tests were conducted considering the various environmental state conditions o fthe radioactivity and the heat generation caused by the neutron exposure, gamma-ray exposure and massiv econcrete hydration in the biological shielding concrete of the JPDR .

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OUTLINE OF THE JPDR CONCRET E

The biological shielding concrete of the JPDR was constructed with the Portland cement concrete as th emixing strength 35 MPa from the March to November 1962. The mixing proportion is shown in Table 1 .The maximum thickness of the concrete was 3 .0 m and had a lining of 13 mm thick steel plate on th ereactor side surface and epoxy paint finishing on the outer side surface . Cooling pipes in order to reducethe thermal heat by radiation exposures and guide tubes of neutron monitoring were installed in the reacto rside concrete .

Table 1 Mixing proportion

The location of the core samplings were selected from the level of the reactor core as the high irradiationconcrete, and the upper and lower distant level from the reactor core as the low irradiation concrete inorder to clarify the influence of the radiation environment and the concrete placing. Figures 1 and 2 showthe sampling location and the concrete core sample .

Fig. 1 Sampling locationof the concrete core

Fig. 2 Concrete core sampled

EVALUATION OF THE ENVIRONMENTAL CONDITION S

Influencing environmental conditions to the biological shielding concrete are radiation exposure and hea tof hydration in the hardening of the massive concrete after the placing . Figure 3 shows neutron flux andexposure dose distribution in biological shielding concrete at the level of reactor core obtained b ycalculation using computer code ANISN-JR. Neutron exposure dose for concrete samples used for th estrength tests were evaluated by the calculation of the neutron flux and exposure dose distribution to the 2dimensional R-Z cylindrical column model using the computer code DOT 3.5 . Distribution graph of the

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Fast neutron flux (E>0 .11 MeV) and thermal neutron flux (E<1 .85 eV) were obtained by the calculatio nusing the computer code DOT 3 .5 . Neutron flux at the test specimens were decided by the distributio ngraph and the neutron exposure dose for the concrete samples were calculated multiplied with th econverted dose for the rated reactor power operation time of the JPDR. Since the error of the calculatedvalue of the thermal neutron flux from the measured one become larger according with the distance apar tfrom the core center to upper or lower directions the calculation was corrected using the radioactivity o fEu-152 measured . Figure 4 shows comparison of the radioactivity of Eu-152 by the calculation and th emeasurement .

The maximum neutron irradiation dose rate to the biological shielding concrete were estimated that the fas tneutron exposure rate was 1 .11 X 10 E+18 n/cm2 and the thermal neutron exposure rate was 4 .75 X10E+17 n/cm2 at the reactor side concrete of the reactor core level . Figure 5 shows the gamma ray fluxdistribution in the biological shielding concrete obtained by calculation at the level of the JPDR reacto rcore. The maximum gamma dose rate obtained 4 .77 X 10 E+8 Gy by the flux converted to the effectiv edose and multiplied with the total rated reactor power operation hours . Figure 6 shows the calorific valuedistribution that was calculated one dimensional transportation analysis using the computer code ANISN-JR generated by the total neutron and the total gamma ray in the biological shielding concrete at the leve lof reactor core . The calorific value in the biological shielding concrete at the reactor side biologica lshielding concrete was obtained 3 .0 X 10 E-4 w/cm3 by neutron exposure, and 5.68 X 10 E-3 w/cm3 by thetotal of primary and secondary gamma-ray exposure . The contribution to the maximum calorific value inthe biological shielding concrete that was obtained 5 .98 X 10 E-3 w/cm3 by the total neutron and the totalgamma-ray was largely contributed by the gamma-ray .

Fig. 3 Neutron flux and exposure dos edistribution in the biological shieldingconcrete at the level of reactor core

Fig. 4 Activity distribution in thebiological shielding concrete at thelevel of reactor core (Eu-152 )

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a o13- " • Case of 45 M

-SHROUD

-VESSÉL PRESSURE

'' . • -

CORE / LINER

i

f

~

I Iin l ~ ~ ~

I .

1

1

1

1

1

l ~

0

40

80

120

160 200

240

280

320

Fig.5 Distribution of total gamma-ray flux at the level of reactor core

Fig.6 Calorific value distributiongenerated by total neutron and tota lgamma-rays at the level of reactor core

The distributions of the concrete temperature for both case of the hydration heat during cure after concreteplacing and the radiation exposure heat during the operation were calculated . The results were shown in theFigures 7 and 8, respectively. As shown in the figures, biological shielding concrete were therma linfluenced by the heat of hydration in the early stage and heat of radiation exposure during the reacto roperation .

Fig. 7 Temperature distribution of thebiological shielding concrete by the heatof hydration at the placing

Fig. 8 Temperature distribution of thebiological shielding concrete in theoperation

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METHOD AND RESULT OF THE TESTS

Test on the mechanical properties of the irradiated concret e

In the tests on the mechanical properties of the concrete, compressive strength, tensile strength, modulus o fstatic elasticity, and Poisson's ratio were investigated . The test specimens were shaped the core samples inthe size of 8 cm diameter and 16 cm height for the tests of compressive strength, modulus of elasticity an dPoisson's ratio, and in the size of 8cm diameter and 8cm height for the tensile strength tests . The concretespecimens were cured in the water 24 hours before the tests . The modulus of static elasticity wereevaluated by the stress-strain ratio at the one third of the maximum stress on the stress-strain curv eobtained by the compression gauge and the strain gauge in the compressive strength tests . The Poisson' sratio was evaluated from the strain ratio of lateral to longitudinal in the linear strain range .

Figure 9 shows the distribution of the compressive strength along with the depths from the reactor side t oouter side concrete . The compressive strength of the concrete core samples was distributed from the rang eof 29.4 to 53 MPa (average 44 MPa) and the average was 20 % larger than the strength of mixin gproportion. The compressive strength showed the tendency to increase along with the fast neutron fluenceincreased in the range from 1X 10E+13 n/cm2 to 1X 10E+17 n/cm2 when it was looked in the relationshi pbetween the compressive strength and the fast neutron fluence calculated as shown in Figure 10 . Since thecompressive strength was influenced by the over burden of the placing height of the flesh concrete at th econstruction, the compression strength were converted to the concrete placing height at 0 cm .

In the previous study, the compressive strength did not decrease in the range of the radiation dose rate2X 10 E+18 to 2X 10E+19 n/cm2 but when the irradiation dose rate is increased to over the 5X 10E+1 9n/cm2 the compressive strength is decreased significantly, and also the compressive strength of theconcrete have been said to decrease by the Gamma irradiation accumulated approximately over 10E+9 G y(Hilsdorf, H.K. et al .) . In our tests, however, the maximum fast neutron and gamma ray dose rates at th ereactor side concrete are 1 .11X 10E+18 n/cm2 and 4 .77X 10E+18 Gy, respectively and it could not confir mthe tendency .

Figure 11 shows the relationships of the modulus of static elasticity and neutron fluence . The influence ofthe neutron fluence to the modulus of elasticity was not shown as the figure shows . The relationship of themodulus of static elasticity and compressive strength were largely distributed around the AIJ curve of th erelations as shown in the Figure 12 .

Fig. 9 Compressive strength alongwith concrete depth

Fig. 10 Compressive strength distributionalong with fast neutron fluence

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Fig. 11 Modulus of static elasticityalong with fast neutron fluence

Fig. 12 Relationship ofcompressive strength and modulusof static elasticity

Figures 13 and 14 shows the relationship between Poisson’s ratio and fast neutron fluence and th erelationship between tensile strength and fast neutron fluence, respectively, obtained by the strength test o fthe concrete core samples . The influence of the fast neutron fluence to the Poisons ratio and the tensil estrength were not shown in the irradiation range of the test samples .

Fig. 13 Poisson’s ratio along with neutronfluene Fig. 14 Tensile strength alon g

with fast neutron fluence

Test on the chemical properties of the irradiated concrete

In order to investigate what influences were affected to the concrete microstructure components by th eirradiation, the tests of the chemical properties of the concrete core samples were performed . Tests item swere chemical element analysis, X-ray diffraction analysis, scanning electron microscope observation ,porosity measurement, water of crystallization measurement, and differential heat analysis .

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Chemical element analysis : The analyses were performed regarding the nine principal elements of Si, Ti,Fe, Al, Mn, Ca, Mg, Na, and K by the method of the Inductively Coupled Plasma Spectrometry . Theoxidation products of the nine principal elements were matched almost the same between the concretesamples of reactor side and outer side as expected.

X-ray difraction analysis : The analyses were performed using the mortar specimens crushed into thediameter less than 45 that were placed uniformly on the glass plate and the diffraction angles weremeasured. The spacing of the crystal faces were obtained input the refraction angles into the Bragg’ scondition and the crystals of the specimen were identified comparing with the standard samples . Theinfluences of the radiation dose effect to the crystallization in the concrete microstructures were not shownfrom the diffraction pattern .

Scanning electron microscope: The observations were performed on two scanning field for one morta rspecimen using the mortar made by roughly crushed the concrete specimen after vacuum drying in scale sup to 500, 1000, and 3000 times larger by the scanning electron microscope . From the scanning electronmicroscope observation at the pore where hydration crystal growth was observed the specimen of reacto rside showed the large growth of the needle crystal than outer side .

Porosity measurement: The micro-pore size of the mortar extracted from the crushed concrete specime nwere measured in the range of 60 to 99,000 + using porosity gauge by the penetration method ofpressurized mercury into the mortar in the pressure range from 0.9 to 2000 kg/cm2. The micro-porediameter distribution of the reactor side concrete was distributed to the small size region than the outer sid econcrete . The peak of the micro-pore distribution of the both side concretes were shown at the diamete r130-250 .

Bound water measurements and diferential thermal analysis: They were performed using the micro-crushed concrete specimen to the size of 45 . The quantities of the water of crystallization were obtaine dfrom the differences after heated the micro-crashed specimen one hours each at the temperatures of 105 ,400 , 650 and 950 by electro heater. The quantity of the free water measured from the loss of weight atthe heating temperature 105 in the reactor side concrete showed 12% larger than the outer side concrete .The differential thermal analysis results were shown not much difference between the reactor side an douter side concrete .

Comparison tests on the mock-up concrete

The test concrete were exposed the thermal condition by heat of hydration during the massive concreteplacing and the heat of irradiation during the reactor operation . In order to clarify the influences of th eirradiation effect and the environmental thermal effect, reference tests were performed on the physical ,mechanical and chemical properties with the same test items using mock up specimens simulated th eenvironmental thermal conditions that are the heat of hydration and the radiation exposure heat as t obecome the same conditions of the actual concrete samples . The test specimens were made to representfour cases of environmental conditions as shown in the Figure 15 . The test cases represent theenvironmental conditions of the actual concrete that are massive concrete cured in an air as case 1, massiv econcrete cured in an air and radiation exposure as case 2, normal concrete cured in a water as case 3, andnormal concrete cured in an air as case 4 .

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Fig. 15 Curing conditions of the mock-up

Fig. 16 Compressive strength of thetest cases

mock-up concrete tests

The test results showed that the effect of the steel lining brought good cure conditions for the concrete inthe both cases 1 and 2 to increase the compressive strength along with longer the curing term as shown i nthe Figure 16 . While the compressive strength of the mock up concrete without steel lining showed theincreasing rate of the long term strength became slower, especially, it were decreased significantly in thecase 2 by the influences of long term heating .

The scanning electron microscope and the X-ray diffraction analysis test results showed not muc hdifference between the cases 1 and 2, but the porosity test results showed the pore size distribution of theboth cases 1 and 2 were shifted to the larger size in the case of without steel lining while in the case of wit hlining it is not changed. The heat of hardening of massive concrete as represented the curing condition i nthe early stage of the cases 1 and 2 were increased the strength generation at the age of one month as larg eas the same age strength of the case 3 . Then the heat of second stage in the case 2 to represent the conditionof the heat by radiation in the operating stage was affected to the concrete without lining to dry th econcrete slowly and evaporate the water content, coarse the hardening body organization, and increased th etotal volume of the micro pores, that seemed to made the strength decrease at the age of 3 month than a tthe age of one month . The specific gravity of the concrete with steel lining did not change according withthe curing term but the concrete without shield lining decreased according with the curing term increase din the both cases 1 and 2 .

CONCLUSION

The evaluation of the radiation exposure dose effect to the concrete structures in nuclear power plants i simportant factor for the plant life management strategy considering aging of the concrete properties duringplant operations. The study using the actual concrete samples from the JPDR biological shielding concret ewas performed when the plant was decommissioning . The major results of the study is summarized asfollows,

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The maximum fast neutron and gamma ray dose rates at the reactor side of the biological shieldin gconcrete are 1 .11 × 10E+18 n/cm2 and 4 .77 × 10E+18 Gy, respectively, at the level of the reactor cor ewhich is almost the same level as the one after 40 years operation of current commercial nuclear powe rplants .

The biological shielding concrete are categorized two featured environmental conditions influenced to thematerial properties that are 1) reactor side concrete with steel lining, and, 2) outer side concrete withoutsteel lining . These conditions were generated relatively high temperatures of concrete by the heat o fhydration of massive concrete in the early stage and the long-term heat by radiation in the reacto roperation .

The generation of heat within the biological shielding concrete by the hydration, neutron and gamma-rayexposures caused large influences to the concrete properties and it were appeared as the variations of th ecompressive strength, the modulus of static elasticity, and the pore size distributions. Poisson’s ratio wa snot shown the influence by the fast neutron dose rate .

The influences of the radiation exposure to the microstructures of concretes were appeared mainly in th ebehavior of the water contents by the long term heating of radiation exposure and it caused the waterdissipation slowly from the outer side concrete without lining. While, the reactor side concrete showedgood durability for the water dissipation by the steel lining even though the decomposition of wate rcontents by neutron exposure as suggested from the high radioactive tritium generation in the reactor sid econcrete .

The compressive strength distribution of the biological shielding concrete were matched the tendency t oincrease along with the fast neutron fluencies increasing within the test range to 10E+17 n/cm2 . Thenegative effects of the radiation dose to decrease the compressive strength of the concrete were no tappeared within the dose range of the test samples to 10E+17 n/cm2 .

Scanning electron microscope observed the large crystal growth of the ettringite in the reactor sid econcrete . In generally, the hydration of the Portland cement is said that aluminate phase C 3A and gypsumCaSO4-2H2O generate the ettringite C3A-3CaSO4-32H2O in the early stage and then it is changing to themonosulfeto C3A-3CaSO4-12H2O. Large growth of ettrigite is said to swell the concrete organization an dto obstruct the strength generation . The large crystal growth of the ettringite in the reactor side concrete i scontradict with the general tendency. It is suggesting that there are still unknown factors about th einfluences of the radiation exposure to the concrete properties .

ACKNOWLEDGEMENT

This report was summarized the past study reports performed by the research people in the Japan Atomi cResearch Institute and the Kajima Corporation when the JPDR was decommissioning . The participants inthe study were greatly appreciated .

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REFERENCE S

Verrall, S ., et al ., Design concepts to minimize the activation of the biological shield of light-wate rreactors, 1985 .

Hilsdorf, H .K., et al ., The effect of nuclear radiation on the mechanical properties of concrete, ACI SP55 -10 .

Idei, Y., Sukegawa T ., et al ., A study on aging of biological shielding concrete of JPDR (I . General studyplan and evaluation of environmental factors), Proceeding of the Fall Meeting of the Atomic Energ ySociety of Japan, pp .31, Oct., 1989 (In Japanese)

Akutu, Y., Idei, Y., et al ., A study on aging of biological shielding concrete of JPDR (II . Strength testresults), Proceding of the Fall Meeting of the Atomic Energy Society of Japan, pp .31, Oct., 1989 (InJapanese)

Idei, Y., Kamata, Y ., Akutu, Y., Kakizaki, M., et al ., Material properties of the JPDR biological shieldin gconcrete, JAERI-M, 90-205, Japan Atomic Energy Research Institute, Nov ., 1990 (In Japanese)

Konno, T., Suzuki, K., et al ., Study of aging for biological shielding concrete in nuclear facilities (I . Testresult of the concrete strength), Proceeding of the AIJ, pp .177-178, Sept ., 1991 (In Japanese)

Kurioka, H ., Kakizaki, M., et al ., Study of aging for biological shielding concrete in nuclear facilities (II.Test result of the chemical properties), Proceeding of the AIJ, pp .177-178, Sept ., 1991 (In Japanese)

Kakizaki, M., Idei, Y., Sukegawa, T ., et al ., Study on environmental and mechanical properties ofirradiated concrete, J . Struct. Constr. Eng., AIJ, No . 488, 1-10, Oct ., 1996 (In Japanese)

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DETECTION OF REINFORCEMENT CORROSION AND ITS USE FOR SERVICE LIFEASSESSMENT OF CONCRETE STRUCTURE S

by C. Andrade*, I . Martinez*, J. Muïïoz*, J. Rodriguez**, M . Ramirez***Institute of Construction Science “Eduardo Torroja”, CSIC, Madrid, Spai n

** Geotecnia y Cimientos S .A. (Geocisa), Madrid, Spai n

1 INTRODUCTIO N

Corrosion of reinforcement is one of the main durability problems of concrete structures . The corrosion i sinduced by two main factors : the carbonation of the concrete cover and the penetration of chloride sproviding from marine atmosphere or from chemicals in contact with concrete . Carbonation generally aim sinto uniform corrosion of the steel bar while chlorides mainly induce localised corrosion . Both types ofcorrosion are of electrochemical nature .

There is a third type of corrosion named stress corrosion cracking, SCC, whose mechanism is not entirel yelectrochemical, but the mechanical stress co-operates for its development . This last type will not beconsidered in present paper.

Reinforcement corrosion is not a common problem in nuclear power plants due to the limited life of thes einstallations, except in case of cooling towers where frequent corrosion problems have been noticed . It canbe however a key aspect to be taken into account when dealing with extension of power plant service life .It is as well a very relevant aspect in long term storage or repository installations where lives beyond 30 0years are usually targeted.

In present paper, it is described first how to measure reinforcement corrosion in order to obtain th ecorrosion rate of the steel . Then, the effects of the evolution of corrosion are listed, in order to b econsidered as limit states and therefore indicators of repair criteria . Finally, a 3D model is presented on th epossible release of chlorides being part of the low and medium radioactive wastes stored in drums . Thesechlorides may diffuse through the surrounding cement mortar and reach the reinforcement of the concret econtainers used to encapsulate the drums . The model is part of a general one that will include not only th eionic diffusion, but also the corrosion of reinforcements and its evolution . This type of models, althoughtheoretical and simple, will help to understand the long-term performance of concrete structures regardin gthe corrosion of reinforcements .

2 ON-SITE TECHNIQUES FOR CORROSION MEASUREMEN T

2.1 Corrosion Potential and resistivity maps .

Up to the present the main techniques used on-site for appraising corrosion of reinforcements are o felectrochemical nature due to that is the basis of the corrosion process .

Because of its simplicity, the measurement of Ecorr (rest or corrosion potential) is the method mos tfrequently used in field determinations . From these measurements, potential maps are drawn which revealthose zones that are most likely to undergo corrosion in the active state 1 . However, such measurement shave only a qualitative character, which may make data difficult to be interpreted 2 . This is due the potentia lonly informs on the risk of corrosion and not in its actual activity . In addition, the developing of macrocell s

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may as well mislead the deductions because corroding zones polarize the surrounding areas, which mayseem corroding as well, when they are cathodic areas of the macrocell . In spite of which potential mappin gstill has a function to accomplish as a qualitative indication of the general performance and a complemen tof the other on-site techniques .The same that said for the potential can be stated on Resistivity, p, measurements 3 , which sometimes areused jointly with Ecorr mapping . The p values indicate the degree of moisture content of the concrete,which is related to the corrosion rate when the steel is actively corroding, but which may mislead th einterpretation in passive conditions. On figure 1 is represented a risk map of a slab . The risk level has beencalculated by a combination of these two parameters : Ecorr and p .

Figure 1 : Corrosion risk map on a reinforcement slab calculated from the combination of Ecorr andmeasurements .

2.2 Polarization Resistance

2.2.1 Laboratory measurements

The only electrochemical technique with quantitative ability regarding the corrosion rate is the so-calle dPolarization Resistance, Rp

4 . This technique has been extensively used in the laboratory. It is based on theapplication of a small electrical perturbation to the metal by means of a counter and a reference electrode .Providing the electrical signal is uniformly distributed throughout the reinforcement, the AE/ I rati odefines Rp . The corrosion current, Icorr, is inversely proportional to Rp, Icorr= B/Rp where B is a constant . Rp

can be measured by means of D .C. or A.C. techniques 5, both of which have specific features in order t oobtain a reliable corrosion current value in agreement with gravimetric losses .

2.2.2 On site measurements

Direct estimation of True Rp values from AE/ I measurements is usually unfeasible in large real concret estructures. This is because the applied electric signal tends to vanish with distance from the counte relectrode, CE rather than spread uniformly across the working electrode, WE . Therefore, the polarizationby the electric signal is not uniform, and it reaches a certain distance that is named the critical length, Lcrit .Hence, AE/ I measurements on large structures using a small counter electrode provides an apparen tpolarization resistance (R p

app) that differs from the true Rp value depending on the experimentalconditions6 . Thus, if the metal is actively corroding, the current applied from a small CE located on th econcrete surface is ’drained’ very efficiently by the metal and it tends to confine itself on a small surfac earea. Conversely, if the metal is passive and Rp is high, the current applied tends to spread far away (e.g . ,around 50 cm) from the application point . Therefore, the apparent Rp approaches the true Rp for activelycorroding reinforcement, but when the steel is passive, the large distance reached by the current needs aquantitative treatment.

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Modulated confinement of the current (guard ring) method .

There are several ways of accounting for a True R p value, among which the most extended one is the use o fa guard ring6, in order to confine the current in a particular rebar area, as Figure 2 depicts . Themeasurement is made by applying a galvanostatic step, lasting 30-100 seconds, from the central counter .Then, another counter current is applied from the external ring, and this external current is modulated b ymeans of the two reference electrodes called “ring controllers” in order to equilibrate internal and externa lcurrents, which enables a correct confinement, and therefore, calculation of Rp . By means of this electricaldelimitation to a small zone of the polarized area, any localised spot or pit can be first, localised, an dsecond its measurement can be made by minimising the inherent error of Rp . Not all guarded technique sare efficient . Only that using a “Modulated Confinement” controlled by two small sensors for the guar dring control placed between the central auxiliary electrode and the ring, shown in figure 2, is able toefficiently confine the current within a predetermined area . The use of guard rings without this contro lleads into too high values of the Icorr for moderate and low values, and the error introduced in the case ofvery localised pits, is very high.

Figure 2 : Modulated confinement of the current (guard ring) metho d

3 EMBEDDED SENSORS

The introduction of small sensors in the interior of the concrete, usually when placing it on-site is bein gone of the most promising developments in order to monitor the long-term behaviour of the structures . Themost usual, as in the case of non-permanent on-site techniques, is to embed reference electrodes o rresistivity electrodes . They can inform of the presence of moisture and on the evolution of corrosio npotential . Others events that can be monitored are the advance of the carbonation or chloride fronts, th eoxygen availability, temperature, concrete deformations and the corrosion rate .

A particular example of the use of embedded sensors is the case of storage facilities of low and mediumradioactive wastes in El Cabril (Côrdoba)7 . There, a pilot container has been instrumented from 1995 byembedding 27 set of electrodes (Figure 3). The parameters controlled are : temperature, concretedeformation, corrosion potential, resistivity, oxygen availability and corrosion rate . The impact oftemperature on several of the parameters is remarkable, and therefore, care has to be taken whe ninterpreting on-site results .

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Figure 3 : Preparation of the embedded sensors in El Cabri l

3 RANGES OF CORROSION RATE VALUES MEASURED ON-SITE

The experience on real structures7 has confirmed the ranges of values previously recorded in laboratoryexperiments4 .

a0.2 gA/cm2 Negligible

0.2 gA/cm2 < Icorr < 0.5 E A/cm2 Low

0.5 gA/cm2 < Icorr < 1 gA/cm2 Moderate

>1 E.,A/cm2 High

Table 1 : Ranges of corrosion rate and risk levels .

In general, values of corrosion rates higher than 1µA/cm2 are seldom measured while values between 0.1 -1µA/cm2 are the most frequent. When the steel is passive very low values (smaller than 0.05-0 .1µA/cm2)are recorded.

A comparison of on-site Icorr values to electrical resistivity has allowed the authors to also rank th eresistivity ones .

Practical measurements of on-site Icorr in El Cabril (Spain )

In order to control long term performance of concrete containers used for low and medium radioactiv ewaste storage, Enresa is developing surveys of a set of parameters in the real structures . Corrosion rate ofreinforcement is one among the parameters measured in the concrete cells in El Cabril – Cordoba – Spain .Figure 4 shows the results of Icorr measured during several years . The values indicate the passivity o freinforcement as expected.

11 8


Ecorr, WALL A

A1 - A2 + A3 -A4 -A5 -A6

RESISTIVITY, WALL A

dic-94 jun-95 dic-95 jun-96 dic-96 jun-97 dic-97 jun-98 dic-98 jun-99 dic-99 jun-00 dic-00 jun-01 dic-0 1

A1 _A2 - A3 - A4 - A5 _A6

A1 -A2 +A3 - A4 - A5 _A6

-400

-300

200dic-94 jun-95 dic-95 jun-96 dic-96 jun-97 dic-97 jun-98 dic-98 jun-99 dic-99 jun-00 dic-00 jun-01 dic-0 1

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Figure 4 : Results of Ecorr, resistivity and Icorr measured during several years over an internal wall of th econtainer .

4 TRANSFORMATION OF ICORR VALUES INTO CALCULATIONS OF LOSS IN BARCROSS SECTION

Corrosion leads into four main structural consequences : 1)reduction of bar cross section, 2) reduction ofsteel ductility, 3) cracking of concrete cover and, 4) reduction of steel/concrete bond (composite effect) .All these effects occurring in isolation, or simultaneously, will result in a loss in the load bearing capacit yof the structure8 .

The primary information obtained from corrosion measurements is that concerning the loss in cross sectio nof the bar. This parameter informs about all the other effects of the corrosion process . The attackpenetration Px is defined as the loss in diameter as is shows in Figure 4. It is obtained through theexpression :

Px = 0. 0115 · IPEP tp

(1)

Being tp the time in years after corrosion started and 0 .0115 a conversion factor of gA/cm2 into mm/year(for the steel) . This expression implies the need to know when the corrosion has started in order to accountfor tp .

When the corrosion is localised (right part of figure 10), the maximum pit depth is calculated b ymultiplying expression (1) by a factor named a which usually takes a value of 10 . Hence expression (1 )above becomes ,

Ppit= 0 . 0115 •I~PEP • tp•a = 0 . 1 1 5 •I~PEP • tp

(2)

UNIFORMCORROSION

Figure 5 : Residual steel section loss considered for the cases of uniform and localised corrosion .

5 MODEL OF SERVICE LIFE PREDICTION OF REINFORCEMENT CORROSION

In general, low and medium radioactivity waste is disposed in concrete containers . In the case of Spain, theprimer container is a cube of 2x2x2 m with a wall thickness of 15 cm . Two layers of nine drums containingthe radioactive waste mixed with cement (cement matrix) are placed inside this container . Once the deck o fthe container has been set, the space left is completely filled with low porosity mortar minimising th enumber of non-desirable air bulbs that could be formed .

In order to study the service life of these containers, several research programs are developed by Enresa(Spanish Agency for Nuclear waste management) . That concerning the service life of cementitiousmaterials in this type of disposal sites is made with the collaboration of the IETcc in Spain .

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5.1 Physical model

Experimental data has shown that transport of ions, from a macroscale point of view, can be describe dthrough an apparent diffusion coefficient Da as a fickian diffusion process . Therefore the flux can beexpressed through the Fick´s First Law :q = -D

apC(x)

(3)Where q denotes the flux vector of chloride, Da is the apparent diffusion coefficient and C theconcentration . Although two materials with different diffusion properties are modelled (mortar andconcrete), each one is considered as a sub-domain in which D a is constant . Besides, the conservation of thetotal amount of ions implies :aC = D div(q)

(4)

atFrom (3) and (4) we obtain, for a constant diffusion coefficient the governing equation can be written as :

(5)= D AC = D + +

t

ap

ap ax2 ay 2 az2

aC

a2C a 2C a 2 C

5.2 Numerical model

The whole model has been meshed with linear hexahedral elements . In some scenarios some regions havebeen meshed with a coarse mesh and others with a denser one depending on the gradient of the chlorid eflux expected . The three main parts of the container (drums, mortar and container walls) have been meshe dseparately in order to assign different material properties . Thirteen scenarios have been modelled.

Figure 6 : Section of the mesh used for the container

5.3

Predictions

Only the concentration history of some critical points has been plotted . A scheme of their location is shownin figure 7 . Since the bottom face of the container has no mortar protection, a point B located on thesurface that contains the bottom bars (BS) is selected . B is within all the points in BS the one that reache sthe maximum concentration values . The same has been done for the point L but related to the lateralsurface (LS) . Curves in figure 7 show that for a constant surface concentration the level of chlorides i s

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continuously increasing (scenario 1) . On the contrary, preserving the total amount of chlorides that initiall ythe active drums contain the curves show a maximum that in all the cases appears before the first 15 0years. This maximum is approximately 30 % of the maximum attained with the first approach . However,this ratio is for the values at 300 years only 10-15 % .

100

150

20 0

T[years]

Figure 7 : Points monitored in the analysis and concentration histor y0 .18 -

0 .16 -

0 .1 4

0 .1 2

0 . 1

0 .0 8

0 .0 6

0 .04

0 .02

0100

10

200

250

300

0

100 T[years] 20 0T[years

5]

Figure 8 : Bottom and lateral critical points concentration histor y

Both graphics in figure 8 show that the values of the point B are in general higher than those of the point L .Moreover, the decrease of the concentration values is more gradual in the second case. In scenario 13, themore critical one, the final concentration levels are quite similar for both points . However, the bars in thebottom wall of the container are subjected to more aggressive conditions during the whole history .

The start of the corrosion can be considered when the chloride concentration values reach the 0 .4-0 .7% ofthe cement weight . That is to say between 0 .068 and 0 .122 % of the total weight of the concrete for acontent of 400 kg/m3 and a density of 2350 kg/m3. The real surface concentration is variable and difficultto determine, but it can attain 6000 ppm. However, the initial surface concentration that reaches th ethreshold level calculated for a final value of concentration of 0 .20 (scenario 13) is :

CS=0.122%/0 .20=0.61%=6100 ppm

(6)

Results show that a constant surface concentration is a roughly approach that must be verified for th emodel under study. Furthermore, the concentration of chlorides decreases when more real boundaryconditions are imposed. The deterioration process is therefore determined not only by the maximu mconcentration level attained, but also by the whole history of the concentration profiles . New experimentsmust be undertaken to show the behaviour of the bars under those conditions .

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The coupled effects of corrosion in concrete and the mechanical behaviour of the reinforcement are no wunder study. Since the chemical environment is very dependent on the materials, the geometry and time ,the mechanical deterioration must be modelled jointly with the transport phenomena .

6 FINAL COMMENTS

Corrosion of reinforcement can be approximately model and accurately measured on-site . The periodicalcorrosion rate measurements on its monitoring through embedded sensors seems very necessary to asses spresent conditions of concrete structures and is a very useful tool in the case of cooling towers of powerplants . Techniques based in the measurement of Polarization Resistance have been implemented i nportable corrosion rate meters to obtain corrosion rate values, and corrosion-data-loggers are now operativ ein pilot containers to monitor corrosion related parameters .

The in-situ techniques should be complemented by models that although still too oversimplified, may hel pto make predictions of the advance of aggressive fronts towards the reinforcements and to predict ver ylong-term performance .

7 ACKNOWLEDGEMENTS

The authors thank to Enresa the funding provided to develop several of the researches presented in th epaper. They thank as well the firm Geocisa for the results of corrosion rate measured by them in El Cabril .

8 REFERENCES

1.

ASTM C876-91 . "Standard Test Method for Half Cell Potentials of Uncoated Reinforcing Steel nConcrete" .

2.

Elsener, B and B6hni, H. Corrosion Rates of Steel in Concrete, N .S. Berke, V.Chaker and D .Whiting (Eds .), ASTM STP 1065, pp . 143-156. (1990)

3.

Millard, S.G. and Gowers, K .R., "Resistivity assessment of in-situ concrete : the influence ofconductive and resistive surface layers", Proc. Inst. Civil Engrs . Struct. & Bldgs, 94, paper 9876 ,pp.389-396 . (1992)

4.

Andrade, C. and G6nzalez, J.A., "Quantitative measurements of corrosion rate of reinforcing steelsembedded in concrete using polarization resistance measurements", Werkst. Korros ., 29, 51 5(1978) .

5.

Andrade, C., Castelo, V ., Alonso, C. and Gonzâlez, J .A., "The determination of the corrosion rate o fsteel embedded in concrete by the Rp on A .C. Impedance methods," ASTM-STP 906, pp . 43-64.(1986)

6.

Feliû, S . , Gonzâlez, J .A., Feliû, S .Jr ., and Andrade, C ., "Confinement of the electrical signal or in-situ measurement of Polarization Resistance in Reinforced concrete," ACI Mater. J., 87, pp 457 .(1990)

7.

Andrade C; Sagrera J .L; Gonzalez J .A; Jiménez F; Bolano J .A; Zuloaga P. "Corrosion monitoring ofconcrete structures by means of permanent embedded sensors" . Niza. Eurocorr’96

8.

Rodriguez, J ., Ortega, L .M. Garcia, A .M., Johansson, L. Petterson, K. "On-site corrosion ratemeasurements in concrete structures using a device developed under the Eureka Project EU-401-Int .Conference on Concrete Across Borders, Odense, Denmark, vol .I, pp .215-226 .

9.

Rodriguez, J ., Ortega, L .M., Casal, J., Diez, J .M., "Assessing Structural conditions of concret estructures with corroded reinforcements", Conference on concrete Repair, Rehabilitation andprotection, Dundee (U .K.), Edited by R.K. Dhir and M.R. Jones, Published by E&FN Spon in Jun e1996, pp.65-77 .

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124


Improved Detection of Tendon Ducts and Defects in Concrete Structures Using Ultrasonic Imagin g

Authors:

W. Müller, V. Schmitz *)M. Krause, M . Wiggenhauser ** )

Fraunhofer Institut für Zerstdrungsfreie Prüfverfahren (IZFP )Bundesanstalt für Materiaailforschung und –prüfung (BAM )

Introduction

At the beginning of the 90s the general opinion was, that ultrasonic inspection methods using pulse-ech otechnique were not suitable for the inspection of concrete because of the inhomogeneity and the strongscattering behavior of the embedded aggregates . In the meantime the progress in the development of ne wequipment and inspection strategies in connection with ultrasonic imaging techniques turns the pulse-ech otechnique into a powerful tool to solve problems related to concrete materials . These imaging techniques –developed for the inspection of homogenous materials like steel or aluminum – could be adopted to thevery low frequencies needed for concrete inspections .

Since 1994 the BAM and the IZFP cooperate in the field of concrete inspection . The BAM perform smeasurements using laser-vibrometers, probe-arrays, and pitch-and-catch arrangements with two probes .The ultrasonic echoes received are digitized and stored on a computer and evaluated at the IZFP using th eSynthetic Aperture Focusing Technique (SAFT) for 3-dimensional imaging . The results are veryencouraging with respect to detection and positioning of reinforcement structures inside the concrete lik etendon ducts and reinforcement bars as well as defects like compaction defects and voids (honey-combin grepresented by styrodur balls) and to detect and even size notches and natural cracks oriented vertical to th esurface . These inspections are applied to bridges, nonballasted tracks, and foundation slabs . They wereperformed in the laboratory as well as in the field at prestressed concrete bridges . In addition the FederalHighway Research Institute (Bundesanstalt für Stral3enwesen, BAST) organized two round robin trials, oneon test specimen containing artificial defects, the other on a motorway bridge, which had to be replaced .Examples of reconstructed ultrasonic images are presented . Further basic research work on ultrasoni cimaging of concrete structures is carried out within a research group of the German Research Counci l(FOR 384 of Deutsche Forschungsgemeinschaft, DFG) .

1 .

A round robin trial with a test block containing artificial defect s

On the concrete support of railway tracks, vertical cracks have been observed . Those cracks allow water topenetrate which causes a corrosion of the tendon ducts and thus reduces the time of life of the supportin gstructure. In the literature different methods based on time-of-flight evaluation of the ultrasonic pulses arereported. Difficulties arise from particles or water which act as ultrasonic bridges between the crac kborders; in those cases the true penetration depth of the cracks can be much larger than the actual measure ddepth extension . Modern research like /1/, /2/ and /2//3/ is aimed to improve the reliability of crack dept hmeasurements .

During the propagation in concrete material with additives of different aggregate sizes, the ultrasoni cpulses are scattered and change propagation direction into all possible directions . In the case of a crackbetween an ultrasonic transmitter and the receiver probe the time-of-flight of received scattered amplitude smoves towards larger values /4/, /5/ .

* )**)

125


Most of the time-of-flight methods are based on the surface wave and the diffraction of the longitudina lwaves at the crack tip . The principle is shown in fig . 1 . For each probe position of the receiver the time-of-flight will be measured . If a surface crack lies between transmitter and receiver the signal will disappea rand come up again with a step in the time-of-flight curve . The time difference is a function of the longersound path around the tip of the crack and proportional to its depth extension .

Fig. 1 Principle of crack depth measurement with acoustic surface and longitudinal wave s

This principle has been used in an experiment at a 20 cm thick concrete test block which contained a 10 c mdeep crack simulated by a notch. The center frequency of the ultrasonic probe was 100 kHz . If the crack i sfilled with water the crack gets penetrable to the ultrasonic pulses and one would expect that this metho dwill fail . In fig 2 the time-of-flight methods are displayed for both case s

Air-filled crack; calculated depth: 91 mm Water-filled crack ; calculated depth: 22 mm

Fig. 2 Problems arise if the crack is filled with water

It is obvious that particles or water in a surface connected crack cause wrong measurmentsmeasurements ,because only the first arrival of the signal is used for evaluation . With regard to cracks in concreteelements reinforcement bars and uncracked aggregates can act as ultrasonic bridges. Therefore it i simportant to develop a more reliable procedure .

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Ultrasonic Synthetic Aperture Focusing Technique (SAFT) has a great potential to image cracks i nconcrete material . Its algorithm superimposes data obtained by pulse echo at many positions which lead sto the suppression of structural noise and to a more reliable positioning of indications .

Computer-based implementations of such a procedure like SAFT were developed by /6/ for the inspectio nof pressure vessels, a combination with ultrasonic holography for defect sizing and classification by /7/ an da three dimensional version implemented by /8/ .

In conventional ultrasonic testing, a specimen is scanned with a narrow search beam to determine th eposition of an object. The situation in concrete testing is different . Given an average velocity of 4000 m/ sin concrete and a transducer of 40 mm diameter a frequency of 200 kHz leads to a divergency beam of 15° ,a 100 kHz transducer to 31° . Hence the ultrasonic beam is not small enough to find the lateral position o fan object . The SAFT algorithm removes this disadvantage .

The movement of a relative small probe imitates a large transducer by sampling its area at many points .This can be done either by an array of transducers which is electronically scanned or by one or twotransducer which are moved step by step . Such a general arrangement is applied to the problem of imaginga surface connected crack – fig. 3 . In pulse echo the transmitter acts as a receiver too and one has to mov ea single probe across the whole surface . If the receiver is separated from the transmitter, it is possible tokeep the transmitter probe fixed and to move only the receiver to different probe positions or to change th eposition of the transmitter simultaneously . In the following a scanning laser Doppler vibrometer is used a sa scanning ultrasonic receiver /8/ . In fig . 4 several transmitter positions have been selected and at eachtransmitter position a two dimensional aperture on the surface opposite to the transmitter position has beenscanned by a laser vibrometer. The data have been superimposed to achieve best quality in the image .

Using these data the reconstruction calculation by means of 3D-SAFT is performed and results in a threedimensional image of reflected and scattered objects from the inside of the specimen

Fig. 3 Scanning arrangements for acoustic imaging with SAFT

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The SAFT algorithm focuses the received signals to any point of the reconstructed image by coheren tsuperposition; that amplitude of each A-scan which may originate from a given voxel due to ist time o fflight value is calculated and averaged into the voxel. Scattered signals are statistical, true reflections notand therefore reflectors can be imaged with a higher signal-to-noise ratio . Scatterer or indications inconcrete are localized at their geometric positions because they are not projected into a B-scan image wit hthe nominal insonification angle of the probe ; the beam opening angle has been introduced into thealgorithm and takes care of all different angles within the divergent sound field .

In fig. 4 an example of a specimen containing a notch is presented . From a 150 mm deep notch verticalslices of the reconstructed image are shown from x- and y-directions .

Fig. 4 Side View SAFT-images from a 150 mm deep empty notc h

What advantages can we take from this procedure?If the crack is partially filled by particles, . they act as scattering center and are imaged in addition to theother reflecting surfaces of the crack . This case was simulated in a specimen where a notch containt a naggregate which acted as a bridge from one side to the other side of the crack. – The result of the SAFTimage for this specimen is presented in fig . 5 . It shows the B-scans from the 3D .SAFT reconstruction intwo directions . The notch tip and the back-wall could still correctly imaged if one compares with th emethod explained in fig . 1 and fig . 2 . The signal from the aggregate is imaged too. In this special case theimage of the aggregate is not correctly located due to another effect . This is a mode conversion andknowing this effect one can locate the bridge into the correct depth . The result shows, that measurement o fa crack depth is possible under site conditions using the imaging system described . This has been verifiedin addition by experiments at real cracks .

128


s

Fig. 5 Correct depth sizing despite a contact spot in the notch due to SAFT-application

2 .

Practical application of ultrasonic testing on a motorway bridg e

In the following the application of the ultrasonic array method including 3D-SAFT reconstruction i sdemonstrated for the localization of honey combing and compaction faults of a base slab with a thicknes sof 300 mm. The specimen has been produced using concrete with a maximum aggregate size of 32 mm an dcontains twofold layers of reinforcement bars on the top and at the back side . The diameter are 25 mm andthe mesh wide 125 mm. The faults were integrated as grains with one constant diameter and styrodur balls .

Fig . 6 Investigation of a 150 long bridge with a base plate and reinforcement

The transducer template was moved over the surface with a step width of 20 mm . The reconstruction in fig.7 shows a vertical slice from the 3D-SAFT reconstruction . The upper and lower reinforcement laye rperpendicular to the moving direction of the transducer template and the back wall echo are clearlydetected. At x = 880 mm and z = 160 mm the reflection from honeycombing is revealed . This is confirmedby the shading of the back wall . A second compaction fault is seen from the shading of the back wall at th eposition of y = 120 mm without direct echo . The localization of the defects matches with the constructionplane with an uncertainty of several cm only. In addition it was possible to image the second layer of th ereinforcement bars . They could not be detected using the impulse radar technology .

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Fig. 7 3D-SAFT: B-Scan of base plate with reinforcementand two artificial flaws 88 mm x 34 m m

In a further application ultrasonic measurements on a post-tensioned concrete bridge deck have bee nperformed . Fig. 8a depicts a 1 m long section through the duct ; duct and back wall are visible on the leftside . A vertical crack – at the position of x = 150 mm –is the reason that neither the back wall nor the duc tcould be imaged. In fig . 18 b and fig. 18 c other sections of the concrete bridge are shown where anoverlap occurs .

130


Fig. 8 Application of ultrasonic inspection on concrete motor way bridges

The results of a detailed analysis can be summarized :

The acquisition of high frequency ultrasonic data with signal processing by an imaging scheme lik eSAFT allows to present an image where the direction and the concrete coverness thickness can b einterpreted .It has been demonstrated that the array system together with 3D-SAFT reconstruction calculation ca nbe used for the examination of transversal prestressed ducts having a concrete cover of about 100 mm .The system has already been successfully used on site .

b .)

131


3. Acknowledgment s

The 3D-SAFT development was supported by the BMWi (Bundesministerium für Wirtschaft) in the fram ework of reactor safety program . Part of the work was developed in the frame of research initiative „NDE ofConcrete Structures using Acoustic and Electromagnetic Echo Methods“ by research funding of th eDeutsche Forschungsgemeinschaft (German Science Foundation), of the Bundesanstalt für Straf3enwese n(Federal Highway Research Instiutute) and by Deutsche Bahn AG (German Railway Corporation) . Thesesupports are gratefully acknowledged.

4. Literature

/1/ Krause, M., Mielentz, F . und Milman, B . : „Machbarkeitsstudie zur zerstdrungsfreienRisscharakterisierung in Festen Fahrbahnen“ ; Abschlussbericht und Anhang, BAM, Juni 2000

/2/ Mielentz, F ., Milman, B. und Krause, M .: „Machbarkeitsstudie zur zerstdrungsfreie nRisscharakterisierung in Festen Fahrbahnen, Phase 2“; Weiterentwicklung der Verfahren für real eRisse in Festen Fahrbahnen, BAM, Sept . 200 1

/3/ Mielentz, F ., Milman, B., Krause, M. und W. Müller, Zerstdrungsfreie Risscharakterisierung i nBetonbauteilen mit Ultraschall,in : Fachtagung Bauwerksdiagnose - Praktische Anwendunge nZerstdrungsfreier Prüfungen, 25 .-26. Oktober 2001 in Leipzig, DGZfP-Berichtsband auf CD ,Posterbeitrag 24, Berlin (200 1

/4/ Kroggel, O .: “Ril3beschreibung mittels Ultraschallreflexion “ Structural Faults and Repair-97 ;Concrete and Composites . Vol . 2 . Proceedings of Seventh Int . Conf. On Structural Faults and Repair,Edinburgh, GB, 9 th . July, 1997, p . 415-42 1

/5/ Frederick, J.R., Seydel,J .A., and Fairchild, R.C . : “Improved Ultrasonic Nondestructive Testing ofPressure Vessels, NUREG-0007-1 report ; U.S. Nuclear Regulatory Commission, University ofMichigan, Department of Mechanical Engineering, Ann Arbor, Michigan, 197 6

/6/ Müller, W., Schmitz, V ., and Sch~fer, G . : “ Defect Sizing and Classification using HOLOSAFT” ,Proc ., 6th Int .Conf. NDE in the Nuclear Industry, American Society for Metals, Metal Park, Ohio, 153 -158, 1983

/7/ Müller, W . : “Betonprüfung” ; IZFP Report No . 940109-E, Institut für Zerstdrungsfreie Prüfverfahren,Saarbrücken, Germany, 1994

/8/ Krause, M., Müller, W. and H. Wiggenhauser, “Ultrasonic Inspection of Tendon Ducts in Concret eSlabs using 3D-SAFT”, Acoustical Imaging Vol . 23 (1997) pp . 433-439

132


Structural Integrity Evaluation of Kori-1 NPP Steel Containment for the Replacement of Stea mGenerator

Yong-Pyo Suh, Korea Electric Power Research Institute, KORE AJong-Rim Lee, Korea Electric Power Research Institute, KOREA

Yeon-Seok Jeong, Korea Institute of Nuclear Safety, KORE A

ABSTRACT

In order to replace the steam generator at Kori unit 1 NPP, in 1998, a comprehensive program for th estructural safety evaluation of steel containment was performed. Based on the replacement process, th echeck list for inspection was made to do the replacement work efficiently and prevent any mistakes .Overall replacement process was appropriately formulated based on regulatory requirements related t onuclear design and repair . As a result of inspection according to the check list, several problems werefound such as adhesive defect at interface between old and fresh concrete and appropriately corrected .

FEM analysis in order to determine the ultimate load of cylindrical steel shell with opening wa sperformed using ABAQUS . Stress concentration and second order deformation due to crane load wasinvestigated through FEM analysis considering inelastic large deformation . It was verified that the currentapproximation analysis using combined elastic buckling criteria gives conservative results . The analyticalresult has shown that the structure follows elastic load-deflection behavior under the given crane loa dcondition with safety factor of 10 .3 . The result of this study will give useful information to the replacemen tof the steam generator of a nuclear power plant .

1. INTRODUCTION

In the field of electric power generation, an importance of nuclear power has been increased becaus eof its large portion of electric facilities in Korea . Recently, extension of lifetime in the field of maintenanc eof nuclear power plants has been main concern from economical point of view . Among equipment in plant,the steam generator is one of the most important component that affect the lifetime of nuclear power plant .

KEPCO carried out the replacement of steam generator at Kori-1 nuclear power plant in 1998 . Inorder to upgrade the steam generator to be safer and more stable, KEPCO planned the Kori-1 SGR(SteamGenerator Replacement) project[1,3] . The containment vessel consists of a 32m in diameter and 44.5m inheight cylindrical shell with thickness of 36 .5mm and a spherical cap with thickness of 19mm .

In order to accommodate the steam generator replacement, an about 7m×7m opening hole was madetemporarily as shown in Fig. 1 and filled back after replacement. A built-in polar crane was to be used fo rlifting and transporting the steam generators for the replacement .

To ensure the structural safety of the containment vessel, a comprehensive program for the structura lsafety evaluation of the Kori-1 Nuclear Power Plant was started in 1998 and was closed successfully . Inthe program, the main processes of inspection were defined and detailed check list was developed. Thereare many concerns in the constructing of safe structure. Specially, the stability of steel containment vesse lagainst crane loading was considered as an important factor to guarantee the safety of the whole structure .The non-linear analysis of the steel containment vessel was performed in order to determine the ultimat eload of steel vessel with opening .

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Fig. 1 Containment Vessel Cross Section for Kori-1 SGR Project

2. MAIN PROCESS AND CHECK LIST FOR THE INSPECTIO N

The inspection to containment vessel should be performed in accordance with nuclear regulator yrequirements, such as regulatory guide, and ACI, ASME, and ASTM code, and so on . In order to ensurestructural soundness and functionality of containment vessel, main process for the inspection in this projec twas defined as the following.

1.

The cutting process of concrete shield building and rigging process of concrete block .2.

The cutting process of steel containment vessel and rigging process of steel plate .3.

Damage of containment facilities such as polar crane bracket during rigging process of stea mgenerator .

4.

The welding process of steel containment vessel .5.

Defect identification of welding part and repair process .6.

The process of concrete mix and production .

134


7.

Pressure test of steel containment vessel .8.

Defect identification of opening concrete containment and repair process .

Detailed check list for the inspection was developed to do the replacement work efficiently an dprevent any mistake and attached in appendix A .

(a) Axial Compression

(b) Axial Compressio nonly

with Internal Pressur e

(c) External Pressureonly

Fig. 2 Basic Loading Components of Buckling Stresses for Cylindrical Containment Vessel

3. STABILITY OF CONTAINMENT VESSEL

The primary concern in the Kori-1 SGR project was to investigate the stress distribution around th eopening hole resulting from the crane operation during the period of replacement of steam generators . Thestress resultants from the SAP90 finite element analysis were put into the “buckling criteria” prepare doriginally for the design of Kori-1 nuclear power plant in the previous work[4] .

135


Fig. 2 shows basic loading components of buckling stresses for the cylindrical containment vesse lunder various loading ; (a) the axial compression only, (b) the axial compression with internal pressure, (c )the external pressure only, (d) the torsion only .

Equations for the basic components of buckling stresses were derived by using the theory o fmechanics and supplemented by experimental observations . In application of the basic buckling stres scomponents to the cylindrical containment vessel subjected to the various types of loading, they wer ecombined by two or three. The analysis revealed that the hole at the steel vessel did not decrease thecapacity of the vessel .

However, when a circular cylindrical containment vessel is under the various loading, shell element saround the opening hole may be subjected to more severe stress concentration than the other parts . Sincethe probable failure mode of the shell elements in such a condition has not been clearly defined, th ebuckling criteria may not be directly applicable to this case, in which an opening hole exists, as was to thecontainment vessel of Kori-1 SGR, in which an opening hole is not existing . It should be noted that thebuckling criteria were originally derived for the containment vessel in service in which there may be som einternal pressure existing due to the operation of steam generators . However, in Kori-1 SGR project atemporary opening hole is existing in the containment vessel that means any internal pressure cannot b egenerated during the period of replacement. Accordingly, it should be recognized that the loadingcondition over the containment vessel with and without an opening hole is very much different. Theelement around the opening hole may be subjected to rather yielding than buckling .

4. ULTIMATE STRESS ANALYSIS OF STEEL CONTAINMENT VESSEL

Because the element around the opening hole of a containment vessel may be subjected to rathe ryielding than buckling, nonlinear analysis considering plastic deformation was required . So, it is performedultimate stress analysis, considering geometric and material non-linearity, of cylindrical containment vesse lusing finite element program ABAQUS in order to estimate the stability of containment vessel with a7m×7m opening subjected to crane loading at both edges of polar crane . To evaluate the effect of stiffeneraround the opening, the cylindrical containment vessel with or without stiffener was analyzed .

Fig. 3 3-D Finite Element Mode l

136


Fig. 3 shows the 3-D analytical model for investigating the ultimate structural behavior o fcontainment vessel . The load applied to finite element model was increased and the stress an ddisplacement was calculated. Fig. 4 shows Von Mises stress distribution around the cylindrica lcontainment vessel subjected to ultimate loading without stiffener and Fig. 5 shows Von Mises stres sdistribution around the cylindrical containment vessel subjected to ultimate loading with stiffener .

Fig. 4 depicts the relation of crane loading vs . displacement at node 2431 (which is located the polarcrane in Fig. 3) . X axis in Fig. 4 shows the displacement at node 2431 and Y axis is the magnificatio nfactor of crane loading .

In the Fig. 5 and Fig. 6, it is shown that an installing stiffener around opening is effective to preventdeformation and stress concentration of the opening . In the Fig. 7, when the value of magnification facto rof crane loading is within 1, cylindrical containment vessel behaves linear elastic . The maximum value ofVon Mises stress is 2 .47×103 t/m2 at area around opening with stiffener and is 9% value of Von Mises' syield criteria 2 .67×104 t/m2

25

20

0 .0 4

Vertical Disp . of Node 2431 (unit : M )

Fig. 4 The Relation of Crane Loading vs . Displacement at Node 243 1

Fig. 5

Von Mises Stress Distribution

Fig. 6 Von Mises Stress Distribution

subjected to Ultimate Loading without

subjected to Ultimate Loading with

137


Stiffener

Stiffener

5. CONCLUSION

In order to replace the steam generator at Kori unit 1 NPP, in 1998, a comprehensive program for th estructural safety evaluation of steel containment was performed. Based on the replacement process, th echeck list for inspection was made to do the replacement work efficiently and prevent any mistakes .Overall replacement process was appropriately formulated based on regulatory requirements related t onuclear design and repair . As a result of inspection according to the check list, several problems werefound such as adhesive defect at interface between old and fresh concrete and appropriately corrected .

It was recognized that the direct application of the buckling criteria[2] for the Kori-1 SGR seems tobe inappropriate . Therefore, in order to estimate the ultimate capacity of containment vessel with a nopening, ultimate stress analysis was performed in consideration of geometric and material non-linearity o fcylindrical containment vessel against crane loading at both edges of polar crane . As a result, it is foundthat when the cylindrical containment vessel is in service loading state, the vessel showed linear elasti cbehavior and service load is about 9% of ultimate load capacity . Therefore, it is concluded thatstrengthening design due to opening is not needed .

REFERENCE S

Bechtel-Hyundai, "Calculation Sheet for Kori-1 Steam Generator Replacement,"

Job No.23213 ,1996 .Lee, A.J.H., "Buckling of Shells under Various Types of Loading," Technical

Report, GilbertAssociates, Inc., 1971 .Westinghouse Electric International Company, "Containment Vessel Design Report for KoriNuclear Power Station - Unit #1," 1972 .Lee, K.J., et al ., “Structural Integrity Evaluation of Kori-1 NPP Steel Containment for theReplacement of Steam Generator”, SMIRT 15, Seoul, Korea, 1999 .

138


APPENDIX A

Check List for the Inspection in the Replacement of Kori-1 Steam Generato r

Object Check Point sinstallation of temporary opening o nthe wall of steel containment vessel

1 . corresponding work procedure and design drawings2. cutting process and condition of temporary opening

after cuttinginstallation of temporary opening o nthe wall

of the

concrete

shieldbuilding

1 . whether the cutting work of concrete block isperformed under the appropriate operating mode ornot.

2 . whether rigging of concrete block damages the stee lcontainment vessel and concrete shield building ornot.

recovery of temporary opening onthe wall of the steel containmentvessel

1 . whether the surface treatment and cleaning ofwelding part are performed appropriately or not.

2. welding procedure and qualification of weldingpersonnel and welding material

3 .

defect

identification

of

welding

part

andappropriateness of repair work

4. structural integrity evaluation of steel containment(pressure test)

recovery of temporary opening onthe wall

of the

concrete

shieldbuilding(rebar work)

1 . corresponding work procedure and design drawings2. cutting process and condition of temporary opening

after cutting3. appropriateness of mechanical rebar splice method4. appropriateness of rebar welding method5. whether rebars are installed appropriately or no t


of the

concrete

shieldbuilding(concrete mix and production)

1 . whether the physical and chemical characteristics o fcement and admixture are in compliance with th ecorresponding technical guidelines or not

2. whether the certification and confirmation tests offine

and

coarse

aggregate

are

performedappropriately and the those results are acceptable o rnot

3. certification of testing personnel and appropriatenessof testing equipment

4. appropriateness of administrating test5 . appropriateness of concrete mix design6 .

certification of batch plant and inspection andcorrection of measuring devices

7. whether the row material and temperature control o fhot weather concrete are appropriate or not


of the

concrete

shieldbuilding(concrete work and curing)

1 . appropriateness of chipping work, cleaning, wettin gin the contact surface between old and fres hconcrete

2. appropriateness of form work design and installation

139


Object Check Point s3. whether the mortar thickness at the contact surfac e

between old and fresh concrete is conformed to therequirement of technical guidelines or no t

4. whether the concrete pouring work is performed i ncompliance with the temperature requirement of ho twhether concrete or not

5. whether the concrete quality assurance is performedin compliance with technical guidelines and tha tresult is appropriate or not.

6 . appropriateness of the pouring and vibration of freshconcrete

7. occurrence of cold joint in hardened concrete8. appropriateness of form removal based on the

technical guidelines9. appropriateness of curing method10. occurrence of defects such as honeycombing an d

voids etc .11 . appropriateness of concrete repair material and

method12. appropriateness of disposition method for the non

conformance things and those result sstructural integrity of containmentinternal structures

1 . whether

rigging of steam generator damages theoperating floor and steel containment and concret eshield building or not

2. whether

rigging of steam generator damages thepolar crane bracket or not

140


New Methods for the Reconstruction of Safety Compartments of Nuclear Power Plant s

Dieter Busch, RWE Solutions AG, Essen, German yProf. Dr . H.-D. Kdpper, Zerna, Kdpper & Partner, Ingenieurgesellschaft für Bautechnik,Bochum, GermanyPeter Holdt, Zerna, Kdpper & Partner, Ingenieurgesellschaft für Bautechnik ,Bochum, Germany

Safety Compartments – Challenges and Construction Specialitie s

Safety compartments are essential structures for the operation of nuclear power plants . Due to theirimportance these concrete structures have to be carefully observed . Certain special requirements must b efulfilled in the construction phase of the structures . When these requirements are not fulfilled the chance o fdamages occurring is great . Simultaneously these structures demonstrate a series of distinctiv econstructional features :

Picture 1 : Total view of a safety compartment whilst reconstructio n

141


A safety compartment is a symmetrical structure with a cylindrical base and crowned with a half-dome(Pic . 1) . The wall thickness varies up to 2 meters in width . Furthermore the diameter of the reinforcementis very large. Due to constructional changes, the concrete cover can vary and in some instances very small .Simultaneously the high number of and large diameter of stirrups present near the concrete surface are als oa cause of the damages .

Found mainly in older structures, the distribution of reinforcement for the reduction in surface cracking i smissing, and therefore cracking does occur .

Damages to be restore d

From the inspections of the concrete surfaces, the following was discovered :

The carbonization depth was up to 15 mm and in some areas the concrete showed detoriation . (Pic . 2)Due to the weather conditions, rain water containing CO 2 and the influence of wind, the concrete surfac ebecame washed out . Additional, found mainly on the top of the safety compartments, microorganism wer eobserved. These organisms have started to produce acids, which added to the wash out of the surface .

Picture 2: Concrete surface of a safety compartment with damages and detoriations

Of greater importance are the surface cracks present . These cracks have reached deep into the concrete ,some of which have reached the reinforcement . These cracks have occurred either due to temperatur e

142


changes that were not taken into account during the construction design, or even from shrinkage crackin goccurring shortly after the completion of construction .

Demands for upgrade

At none of the inspected safety compartments were reconstruction measurements necessary . All the workscarried were to minimize future repairs and to keep the state of art . There was at no time any doubt for thesafety of the concrete structure or the power plant . Furthermore, to reassure the public, nuclear powe rplants should present a perfect optical appearance .

The main objective of the works at the surface of the safety compartments was to stop any furthe rcarbonization . Simultaneously the further development of cracking had to be stopped and existing crack shad to be closed.

Concepts for restoration

The development of the restoration concept was based on the data and experiences gained from the repai rof other concrete surfaces such as cooling towers, stacks, foundation walls, bridges, etc . Their evaluationas well as the first tests on the safety compartments showed that stiff repair systems even with flexibl ecoatings would not achieve a satisfactory solution . Therefore the aspect of crack bridging and weatherprotection was addressed in two steps, simultaneously looking for an architectural satisfactory solution .This concept with its optimal results was only possible due to the new developments of repair and coatingmaterials achieved in the recent years . Table 1 shows the different measures taken in the last twelve year son German concrete safety departments .

Table 1 : Development of repair systems for concrete surfaces in the last twelve year s

The following concept resulted from these experiences :

- Preparation of the concrete by detecting and removing all loose areas by hand and with the use of ahammer

- Blasting of the complete concrete structur e- Preparation of all corroded reinforcement and application of a protective coating .- Injection of cracks when possible, i .e . the wide cracks- Reprofiling of the damaged and deteriorated areas with a PCC Mortar- Closing of pores with a stiff mortar- Application of a mineral based crack-bridging coatin g- Final application with an elastic, pigmented top coat based of pure acrylic resi n

An area of concern was at the edges of the hollow areas, most of which had occurred due to architectura lreasons during the construction. As shown in Picture 3 – at the edge there is the danger that the coverag emay become too thin while on the inside cracks could occur due to too much material being place d

143

Picture 3: Cross – section of a fluting, showing minimal and excess material thicknes s

In former times with the materials available, it was not possible to solve these problems and so hollow swere completely filled with a PCC mortar before applying a crack-bridging mortar . At the “minimumrestored areas” new materials were used. These new materials were able to take care of the problems . Noneof the different techniques used showed a lack in the optical appearance .

144


For all the materials that were developed and tested severely, it was of great importance that the material sremain flexible even at temperatures of minus 20 0 C, as well as keeping at least 50% of its flexibility evenafter 20 years . These demanding requirements were tested on existing safety compartments . The result sshowed crack movements only due to temperature changes .

145


Scaffolding s

An important device for undertaking the repair work is the use scaffolding . In the past a scaffolding wa sconstructed around the whole compartment . This was not only expensive and ineffective, but also did notcomply to the existing safety standards within a German nuclear power plant . These scaffoldings had to b eguarded to ensure that only restricted personnel were allowed to step on the scaffolding . Another problemof this scaffolding construction was the many anchor points into the concrete surface . These anchor point shad to worked on after the scaffold was removed .

Therefore a movable scaffolding/platform was developed, which is fixed to one huge anchor bolt on th every top of the dome. This scaffolding/platform is circulating on wheels around the dome . This systemworked well with the old protection systems coating the concrete . But when the flexible protective system swere used, the forces of the wheels started to damage the applied mortar, as shown in picture 5 .

Picture 5: Damages on the repair system caused by high load on scaffolding wheel s

Therefore a new light weight cradle-type scaffolding was designed, as shown in picture 6 . This lighterscaffolding was used without any problems on the last safety compartments . The light weight scaffoldingcould move in all directions more easily, and like the former model was also fixed to an anchor bolt on th etop .

146


Picture 6 : New designed light weight scaffolding/platform

Concrete protection with crack – bridging system s

The repair of the concrete surface is started, as is the norm with all other concrete structures, by blastin gand cleaning the concrete and the all open reinforcement. The reinforcement has then a protective coatingapplied, and then the concrete is re-profiled . What is unique about the reported system here is the use of acrack–bridging system for the first time. Due to the costs and time it was agreed to use a fine mortar toclose all the pores and small caverns in the concrete surface before applying the flexible system .

Through experience it was observed that leaving out the scratch mortar would result in damaging th ebridging layers . This is due to the lack of thickness or small holes, as shown in picture 7 .

147


Picture 7: Surface of the crack – overbridging material with pores and hollow s

The optimal application method for crack–bridging, is to wet spray apply it onto the surface . By optimizingevery application step it is possible to achieve the required thickness of the layer and to achieve a surfac ewhich can accepted the final coating .

An important area of concern during the application is to pay attention to the change of climate, especiall ythe change in humidity. If these conditions are not regarded carefully, “Calcium-hydroxide / Calcium -carbonate” may develop, which will influences the color and physical properties .

Picture 8 : Perfect surface of the crack overbridging surface protection system

148


Quality Contro l

Most important, beside the use of proper materials and correct techniques is the installation of a qualit ycontrol system from the very beginning . Before the start up of works a quality handbook was created bythe contractor, the owner of the plant and the supervising engineer . This quality handbook contains all thequality checks and the production documentation. Its contents also include the various times andspecifications that the contractor has to control during the repair .

It was also agreed upon how often the supervising engineer would make his quality control checks . Andthat before the next stage of repair commenced the supervising engineer had given his approval . Thereason for this is that the preparation priming coat can not be checked after the coating later .

One of the most important measurements is the determination of the film thickness, for example with theuse of a penetrometer, but other methods without destroying the coating are better . The main goal for therepeated controls, for which the costs were calculated in advance and paid separately, was to reach aconstant and optimal thickness of the crack–bridging and the final layer of coating . If these layers are toothick possible problems may occur in the future .

Picture 9 : Partial View of a restored concrete surface of a safety compartment

Future Aspects

There are no problems with the applied coatings at present . The materials and techniques used on thesafety compartments can be used of course on other concrete structures such as cooling towers and stacks .One problem in Germany has still not been solved . As shown in picture 10, the pigeons are attracted tothese domes, and since their droppings are quite aggressive . Therefore the coating in the pole area has to b echecked carefully in the future .

149


Picture 10 : Completely restored surface, biological attack caused by birds residue s

150


E. LIST OF PARTICIPANTS

CANADAMr. POPOVICManager of Civil EngineeringAECL

Tel: +1 (905) 823 9060 X213 5Fax: +1 (905) 855 947 0E-mail: popovica@aecl .ca

2251 Speakman DriveMISSISSAUGA, Ont L5K 1B2

Mr. Claude SENIMattec Engineering Ltd.William Carson Crescent 217-21 8

North YorkOntario M2P 2G6

Tel: +1 (416) 224 575 1Fax: +1 (416)224 575 1E-mail: [email protected]

BELGIU MMr. Luc DE MARNEFFE Tel: +32 2 773 81 48Principal Engineer Fax: +32 2 773 89 7 0TRACTEBEL Eml: luc .demarneffe@tractebel .be7, Avenue Arian eB-1200 BRUSSEL S

Mr. Roland LASUDRY Tel:TRACTEBEL Fax:7, Avenue Ariane Eml: roland .lasudry@tractebel .beB-1200 BRUXELLES

CZECH REPUBLICMr. Jan MALY Tel: +420 2 41006 EXT 420Energoprojekt Praha a.s . Fax: +420 2 41006 409Vyskocilova 3 Eml: [email protected]. Box 158140 21 Prague 4

Mr. Ladislav PECINKA Tel: +420 2 2094 11 12Senior Research Worker Fax: +420 2 2094 051 9Division of Integrity and technical Engineering Eml: [email protected] RezVltavska 225068 REZ

Mr. Jan STEPAN Tel: +420 2 4100642 1Energoprojekt Praha a.s . Fax: +420 2 4100640 9Vyskocilova 3 Eml: stepan@egp .czP.O. Box 158140 21 Prague 4

15 1


Mr. Petr STEPANEK Tel: +420 5 4114 620 5Associate Professor Fax: +420 5 4321 210 6Brno University of Technology Eml: stepanek.p@fce .vutbr .czFaculty of Civil EngineeringDepartment of Concrete and Masonry Structure sUdolni 53 - 602 00 Brn o

FINLANDDr. Pentti E. VARPASUOFortum Engineering Ltd .POB 10, 00048 Fortum,Rajatorpantie 8, VantaaFIN-0101 9

Erkki VESIKARILic . Sc . (Tech.), Senior Research ScientistVTT Technical Research Centre of Finlan dKemistintie 3, Espo oP .O. Box 1805, FIN-02044 VTT, Finland

Tel: +358 10 45 3222 3Fax: +358 10 43 32022Eml: pentti [email protected]

Tel: +358 9 456 6922Fax: +358 9 456 7003E-Mail : erkki [email protected]

FRANCEMr. TOURET Tel: +33 4 72 82 71 9 4EDF-Septen Fax: +33 4 72 82 77 0 7Basic Design Dept . E-mail : [email protected] and Construction Division

Tel: +33 (0)1 4654 8028

12-14 Dutriévoz Avenue69628 VILLEURBANNE CEDEX

Mr. Jean Mathieu RAMBACHDES/SAMS CE FAR Fax: +33 (0)1 4746 101 4CEA/IPSN Eml: [email protected] Avenue du General-Lecler cB .P . 692265 Fontenay aux Rose s

Mr. Olivier STRICH Tel: +33 (0)1 46 54 93 28IPSN Fax: +33 (0)1 47 46 10 1 4Département d'Evaluation de Sûreté Eml: olivier.strich@ipsn .frB .P . 692265 Fontenay aux rose s

GERMANYMr. Dieter BUSCH Tel: +49 + 49 201 12 24 47 6Junior Assistant Manager Fax: +49 + 49 201 12 22 48 6RWE Energie AG Bereich Bau Eml: dieter .busch@rweplus .comKruppstrasse 545117 ESSEN

152


Mr. Peter HOLDT

Tel : +49 234 9204 18 5Project Managaer

Fax: +49 234 9204 15 0Zerna, Kdpper und Partner

E-mail : hol@zkp .deIndustriestrasse 2744892 Bochum

Mr. Christoph NIKLASCH

Tel : +49 721 608 227 5Research Assistant

Fax: +49 721 608 2265University of Karlsruhe

Eml: christoph .niklasch@ifmb .uni-karlsruhe .deInstitut für MassivbauP.O. Box 6980D-76128 Karlsruhe

Dr. Rüdiger MEISWINKEL

Tel: +49 0511 439 290 6E.ON Kernkraft GmbH Zentrale

Fax: +49 0511 439 4144Nuclear Systems & Components Division

E-mail :[email protected] 5D-30457 Hannover

Dr. Volker SCHMITZ

Tel : +49 681 9 302 387 0Head of Department Quantitative NDE

Fax: +49 681 9 302 593 0Fraunhofer IzfP

Eml: schmitz@izfp .fhg.deUniversitaet, Geb 3 7D-66123 Saarbruecken

Dr. Friedhelm STANGENBERG

Tel : +49 + 49 (0) 234 961301 2Stangenberg und Partners, Consultants

Fax: +49 + 49 (0) 234 9613048Ingenieur-GmbH

Eml: [email protected] 47D-44787 BOCHUM

Dr. Herbert WIGGENHAUSER

Tel : +49 (0) 30 8104 1440Bundesanstalt für Materialforschung

Fax: +49 (0) 30 8104 1447und -prüfung

Eml: herbert .wiggenhauser@bam .deDivision IV.4 Non-Destructive Damag eAssessment and Environmental Measurement Method sUnter den Eichen 8 7D-12205 BERLIN

Herr Rüdiger DANISCH

Tel: +49 (0) 9131 189 342 6FRAMATOME ANP GmbH

Fax: +49 (0) 9231 189 759 9NDA2

E-Mail : ruediger.danisch@framatome-ANP .deP.O. Box 322091050 Erlangen

Herr Andreas KOCHANMC Bauchemie Bottrop GmbHAm Kruppwald 2-846238 Bottrop

153


HUNGARYMrs. Katalin GYARMARTH YEngineerNuclear Power Plant Co .PAKS

Tel : +36 75 50 88 6 3Fax: +36 75 50 65 3 5Eml: gyarmathy@npp .hu

Prof. Dr. Peter LENKEI Tel : +36 72 224 268 ext 7237Professor of Structural Engineering Fax: +36 72 214 26 8PÉCS University, College of Engineering Eml: lenkeip@witch .pmmf.huH-7624 PECSBoszorkany U.2

Mr. Csaba NYARADI Tel: +36 7550 705 4Systems Technologist Fax: +36 7550 7334Nuclear Power Plant Paks Eml: [email protected] PAKS P .O.B. 7 1

Mr. Oliver KAKASY Tel: +36 (75) 508 93 9Resident Inspector Fax: +36 (75) 311 47 1Hungarian Atomic Energy Authority E-mail: [email protected] .huNuclear Safety DirectorateH-1539 Budapest 11 4P.O.B. 676

ITALYMr. Alberto TAGLIONI Tel : +39 06 30483 3628ENEA Fax: +39 06 3048 630 8Via Anguillarese 301 Eml: [email protected] ROMA

Dr. Lamberto D’ANDREA Tel: +39 (0) 6 83 04 03 5 0SOGIN S.p .A . Fax: +39 (0) 6 83 04 04 74Via Torino 6 E-mail: [email protected] Rome

JAPANMr. Takaaki KONNO Tel: +81 3 3581 9842Secretariat of Nuclear Safety Commission Fax: +81 3 3581 9836Cabinet Office Eml: tkonno@op .cao .go .jpTechnical Counsellor3-1-1 Kasumigaseki, Chiyoda-kuTokyo 100-8970

154


KOREA (REPUBLIC OF )Dr. Yun Suk CHUN GPrincipal Research EngineerKorea Institute of Nuclear Safety19 Guseong,YuseongTaejon 305-33 8

Mr. Yun-Suk CHUNGResearch Project ManagerKorea Institute of Nuclear Safety19 Guseong, Yusung,Taejon 305-33 8

Dr. Jeong-Moon SEOProject ManagerKorea Atomic Energy Research Institut eP.O. Box 105YusongTaejeon 305-60 0

Mr. Yong-Pyo SU HSenior Member of Technical StaffKorea Electric Power Research Institute103-16 Munji-DongYusong-GuTaejon 305-380

SLOVAKIAMr. Juraj NOZDROVICKYProject ManagerVUEZ, a.s .Sv. Michala 4,P.O. Box 153934 80 LEVIC E

Mr. Milan PRANDORFYProject ManagerVUEZ, a.s .Sv. Michala 4P.O. Box 153934 80 Levice

Tel : +82 42 868 053 3Fax: +82 42 861 9945Eml: k063cys@kins .re .kr

Tel: +82 42 868 0533Fax: +82 42 861 9945Eml: k063cys@kins .re .kr

Tel : +82 42 868 839 1Fax: +82 42 868 8374Eml: jmseo@kaeri .re .kr

Tel: +82 42 865 579 1Fax: +82 42 865 5504E-mail: ypsuh@kepri .re .kr

Tel : +421 36631 366 5Fax : +421 36631 366 3Eml: mechanika@pobox .sk

Tel : +421 3663 13665Fax : +421 3663 13663Eml : [email protected]

155


SLOVENIAMr. Lojze BEVC

Tel : +386 1 28 04 48 7Head of Structural Department

Fax: +386 1 28 04 484Slovenian National Building and

Eml: lojze .bevc@zag .ziCivil Engineering InstituteDimiceva 12,St-1000 LJUBLJANA

Mr. Bozo KOGOVSEK

Tel: +386 1 477 62 03Project Manager

Fax: +386 1 251 05 27IBE Consulting Engineers

E-mail : bozo [email protected] iHajdriho va ul 41000 Ljubljana

SPAIN

Mrs. Dora LLANOS

Tel: +34 942 245 100Head of Civil Structure Section

Fax: +34 942 245 123NUCLENOR, S .A .

Eml: dora .llanos@nuclenor .esCalle Hernàn Cortés 2639003 Santander

Mr. Jesus GARCIA ROCASOLANO

Tel: +34 91 659 8726TECNATOM

Fax: +34 91 659 86 77Avenida Montes de Oca No . 1

E-mail : jroca@tecnatom .es28709 San Sebastian de los Reye sMADRID

Mr. Juan SABATER

Tel : +34 97781870 0Civil Engineer

Fax: +34977818720C.N. ASCÔ / C.N . VANDELLÔS II

Eml: jsabater@anacnv .comct . n.340 km. 1123Apartado de Correos 4 843890 L'Hopitalet de l'Infant

Mrs . MARTINEZ SIERRA

Tel : +34 91 3020440Instituto Eduardo Torroja

Fax: +34 91 3020700Spanish Research Council

E-Mail : [email protected] .esSerrabi Galvache StreetES 28033 Madrid, Spain

SWEDENDr. Behnaz AGHILI

Tel : +46 8 698 8692Swedish Nuclear Power Inspectorate

Fax: +46 8 661 908 6Klarabergsviadukten 90

Eml: behnaz .aghili@ski .seStockhol mSE-106 58

156


Mr. Patrick ANDERSONDivision of Structural Engineerin gLund UniversityP.O. Baox 1 8SE-22100 LUND

Mr. Gabriel BARSLIVOSwedish Nucler Power InspectorateDepartment of Structural Integrity(SKI)S 10658 Stockholm

Mr. Jonas BERGFORSProject ManagerOskarshamn Nuclear Power PlantSE-572 83 Oskarshamn

Mr. Jan GUSTAVSSONManager Y2K Projec tRinghals Nuclear Power PlantVattanfall ABRinghal sS-430 22 VAROBACKA

Mr. Thomas ROT HDepartment of Structural EngineeringRoyal Institute of Technology (KTH )SE 10044 Stockholm, Sweden

Mr. Thomas VIBERGProject ManagerOskarshamn Nuclear Power PlantSE-572 83 OSKARSHAMN

SWITZERLANDMr. Jean-Baptiste DOMAGEMonitoring Development Manage rVSL Schweiz AGIndustriestrasse 14CH-4553 Subingen

Tel : +46 46 2984 3Fax: +46 46 24212Eml: [email protected] .se

Tel : +46 8 698 866 0Fax: +46 8 661 908 6Eml: gabriel .barslivo@ski .se

Tel: +46 491 78 79 4 3Fax: +46 491 78 60 3 8Eml: jonas .bergfors@okg .sydkraft .se

Tel : +46 340 66 79 5 0Fax: +46 340 66 83 8 9Eml: jan .gustavsson@ringhals .se

Tel : +46 (0)8 790 813 6Fax: +46 (0)8 21 69 4 9Eml: [email protected] .se

Tel : +46 0491 786 20 5Fax: +46 0491 786 038Eml: thomas.viberg@okg .sydkraft .se

Tel : +41 32 613 30 7 2Fax: +41 32 613 30 7 5Eml: jbdomage@vsl-schweiz .ch

UNITED KINGDOMMr. Robin BALDWINMott MacDonald Limite dMaterials Technology UnitBristol Office - 0117 906 9529Prince House, 43-51 Prince Stree tBristol BS1 4PS

Tel : +44 (0)181 774 2000Fax: +44 (0)181 681 5706Eml: [email protected]

157


Dr. Tony MCNULTYNII, HSENuclear Safety directorateSt Peter’s HouseBalliol RoadBootle, Merseyside L20 3JZ

Dr. Leslie M . SMITHSenior Civil EngineerBritish Energy Generation (UK) Ltd3 Redwood Crescent, Peel ParkEast Kilbride G74 5PRGLASGOW

Tel: +44 151 951 362 4Fax: +44 151 951 4163Eml: tony .mcnulty@hse .gsi .gov.uk

Tel : +44 (13552) 6238 5Fax: +44 (13552) 6245 9Eml: les .smith@british-energy .com

UNITED STATES OF AMERICADr. Dan NAUS Tel: +1 865 574 065 7Oak Ridge National Laboratory Fax: +1 865 574 2032PV Tech Sect, Eng Tech Div E-mail : nausdj@ornl .govP.O. Box 2009, Bldg .9204- 1OAK RIDGETN 37831-8056

Dr. James F . COSTELLO Tel: +1 (301)415-600 4Office of Research Fax: +1 (301)415-5074Division of Engineering Technology E-mail : jfc2@nrc .govMail Stop T10-L 1US Nuclear Regulatory CommissionWashington, DC 20555

International Organisations

International Atomic Energy Agency, ViennaMr. Paolo CONTRI

Tel : +43 1 26000 26426International Atomic

Fax: +43 1 26007Energy Agency

Eml: p [email protected]/NS/NSNI/ES SWagramerstrasse 5P.O. Box 100 A-1400 VIENNA

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Documents

Repair of Concrete Nuclear Power Plant